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1/24/83      ~
UNITED STATES OF AMERICA              cgtg.(Tyc NUCLEAR REGULATORY COMMISSION
                                                                    '83 JAN 24 P3:05 In the Matter of                                        )
                                                            )                    _.
UNITED STATES DEPARTMENT OF ENERGY                      )                  c u n'r
                                                              )
PROJECT MANAGEMENT CORPORATION                            )  Docket No. 50-537
                                                              )
TENNESSEE VALLEY AUTHORITY                                )
                                                              )
(Clinch River Breeder Reactor Plant)                      )
                                                              )
APPLICANTS' PROPOSED PARTIAL INITIAL DECISION George L. Edgar Thomas A. Schmutz Frank K. Peterson Morgan, Lewis & Bockius 1800 M Street, N.W.
Washir_gton, D. C. 20036 Attorneys for Proj ect Management Corporation Leon Silverstrom Warren E. Bergholz, Jr.
Uilliam D. Luck U. S. Department of Energy Office of General Counsel 1000 Independence Avenue, S.W.
Room 6B-256 -- Forrestal Bldg.
Washington, D. C.      20585 Attorneys for United States Department of Energy Herbert S. Sanger, Jr.
Lewis E. Wallace W. Walter LaRoche James F. Burger Edward J. Vigluicci Dated:              January 24, 1983              Tennessee Valley Authority 400 West Summit Hill Drive Knoxville, Tennessee 37902 Attorneys for the Tennessee Valley Authority 8301250477 830124
{DRADOCK05000
 
GLOSSARY ACRS  Advisory Committee on Reactor Safeguards AEA  Atomic Energy Act AEC  United States Atomic Energy Commission A Exh Applicants' Exhibit AFW  Auxiliary Feedwater ATWS  Anticipated Transients Without Scram AW    Applicants' Witness BEIR  Biological Effects of Ionizing Radiation BNFP  Barnwell Nuclear Fuel Plant CCTV  Closed Circuit Television CE    Commonwealth Edison Company CFE  Clandestine Fission Explosive CP    Construction Permit CRBRP Clinch River Breeder Reactor Plant DBA  Design Basis Accident DHRS  Direct Heat Removal Service DOE  United States Department of Energy DOT  United States Department of Transportation DRP  Developmental Reprocessing Plant EBR  Experimental Breeder Reactor EPA  United States Environmental Protection Agency ER    Environmental Report
 
ERDA      United States Energy Research and Development Administration FES      Final Environmental Statement FFTF      Fast Flux Test Facility FMEF      Fuels and Materials Examination Facility FSFES      Final Supplement to Final Environmental Statement FWS        United States Fish and Wildlife Service GAO        United States General Accounting Office HEPA      High Efficiency Particulate Absolute HLW        High Level Waste HTS        Heat Transport System I Exh      Intervenors' Exhibit IHTS      Intermediate Heat Transport System ICRP      International Commission on Radiological Protection INEL      Idaho National Engineering Laboratory IW        Intervenors' Witness JCAE      Joint Committee on Atomic Energy l
LDP        Large Development Plant LEID      Limit of Error on Inventory Differences LLW        Low Level Waste LOF        Loss of Flow LOHS      Loss of Heat Sink LPZ        Low Population Zone LMFBR      Liquid Metal Fast Breeder Reactor l
i
 
LWA  Limited Work Authorization LWR  Light Water Reactor MC&A  Material Control and Accountability MOX  Mixed Oxide MWt  Megawatts Thermal MWe  Megawatts Electrical Na    Sodium NCRP  National Council on Radiation Protection and Measurement NDA  Non-Destructive Assay NEPA  National Environmental Policy Act NRC  Nuclear Regulatory Commission NRTA  Near Real Time Accounting ORGDP Oak Ridge Gaseous Diffusion Plant ORNL  Oak Ridge National Laboratory PAG  Protective Action Guideline l
t  PCD  Population Center Distance PRTS  Primary Heat Transport System PSAR  Preliminary Safety Analysis Report l
PMC  Project Management Corporation poig  pounds per square inch gauge Pu    Plutonium PWR  Pressurized Water Reactor RAPS  Radioactive Argon Processing System RSS  Reactor Shutdown System
 
SACOS      Safeguards Computer Operating System S Exh      Staff's Exhibit SGAHRS    Steam Generator Auxiliary Heat Removal System SNM        Special Nuclear Material SRP        Savannah River Plant SSR        Site Suitability Report SST        Safe Secure Truck SSST      Site Suitability Source Term SW        Staff's Witness TOP        Transient Overpower TR        Transcript Page TRU        Transuranic TSS        Transportation Safeguards System TVA      The Tennessee Valley Authority UNSCEAR  United Nations Scientific Committee on the Effects of Atomic Radiation VEPCO    Virginia Electric Power Company ZPPR      Zero Power Plutonium Reactor
 
TABLE OF CONTENTS Page(s)
I. BACKGROUND....................................                                        1 II.        INTRODUCTION..................................                                    13 III.      OPINION........................................                                    15 A.      Site Suitability Issues...................                                16
: 1. Contentions 1, 2, and 3 (Site Suitability)....................                                16
: 2. Contention 2e)                (Site Suitability Dose Guidelines)......................                                37 B. Environmental Issues......................                                43
: 1. Contentions 2d), 2f), 2g), 2h), 3c) and 3d) (Environmental Effects of Severe Accidents).....................                                44
: 2. Contention 5b) (Risks to Nearby Energy and National Security Facilities Due to CRBRP Accidents)....                                  50
: 3. Contentions 4 and 6b)4) (Safeguards Impacts)..............................                                55
: 4. Contentions 6b)l) and 6b)3) (Fuel                                      62 cycle Impacts)........................
: 5. Contentions Sa) and 7c) (Alternative                                  66 Sites)................................
: 6. Contentions 7a) and b) (Likelihood of Meeting Programmatic Objectives and Alternative Designs)..............                                70
: 7. Contentions lib) and c) (Genetic and Somatic Ef f ects of CRBRP Operation) . . .                            72 1
 
C. Ultimate Legal Issue......................          75
: 1. The Atomic Energy Act Requires That Section 50.10(e) be Available to the            76 CRBRP.................................
4
: 2. The Rulemakin  History of Section 50.10 e)......................          78 IV. CONCLUSION....................................          80 V. ORDER.....................................'....
                                ~
82 VI. FINDINGS OF FACT..............................        F- 1
(  6- 71) - Radiological Site Suitability Issues. . . . . . . . F- 2
                - contentions 1, 2 and 3 (Site Suitability)        F- 2
(  7- 46)
( 47- 60)    - contention 2e)                                    F-54
( 61- 71)    - Ultimate Site suitability Findings                F-64
( 72-274) - Environmental Issues                                  F-67
( 73- 87)    - Contentions 2 and 3 (Environmental)              F-67 F-83
( 88-104)    - contention 5b)
(105-159)    - contentions 4 and 6b)4)                          F-95 (160-189)    - contentions 6b)1) and 6b)3)                      F-126 l (190-209)    - contentions sa) and 7c)                          F-145 l
                - contentions 7a) and b)
F-160 (210-243)
F-183 (244-259)    - contentions lib) and c)
                - Ultimate Environmental Findings                  F-191 (260-274)
VII. CONCLUSIONS OF LAW............................        C- 1
                                          ~
l    APPENDIX A - EXHIBIT LIST............................
APPENDIX B - WITNESS LIST............................
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of                            )
                                              )
UNITED STATES DEPARTMENT OF ENERGY          )
                                              )
PROJECT MANAGEMENT CORPORATION              )    Docket No. 50-537
                                              )
TENNESSEE VALLEY AUTHORITY                  )
                                              )
(Clinch River Breeder Reactor Plant)      ')
APPLICANTS' PROPOSED PARTIAL'' INITIAL DECISION The United States Department of Energy (DOE) and Pro-ject Management Corporation, for themselves and on behalf of the Tennessee Valley Authority (the Applicants), hereby file this Proposed Partial Initial Decision for a Limited Work Authoriza-tion (LWA) pursuant to 10 C.F.R.    { 50.10(e). This Proposed Partial Initial Decision is presented in seven parts:      I) Back-ground, II) Introduction, III) Opinion, IV) Conclusion, V) Order,
                                                        */
VI) Findings of Fact, and VII) Conclusions of Law.-
4 I. BACKGROUND This Partial Initial Decision concerns the application filed with the United States Nuclear Regulatory Commission (the Commission or NRC) by the United States Department of Energy
  */    Enclosed also is a set of Appendices to the Proposed Deci-i      sion consisting of: A) an Exhibit List, and B) a Witness List.
 
(DOE), Project Management Corporation (PMC), and the Tennessee Valley Authority (TVA) for a Construction Permit and Operating License for the Clinch River Breeder Reactor Plant (CRBRP ) . The CRBRP is a Liquid Metal Fast Breeder Reactor (LMFBR) demonstra-tion plant with a rated output of 350 megawatts of electrical power, to be located on the Clinch River in Oak Ridge, Tennessee.
                */
S Exh 8 at 1-1.-    This Partial Initial Decision addresses the portions of the application for a Construction Permit which are necessary for Limited Work Authorization (LWA) findings under 10 C.F.R. $ 50.10(e)(2); namely, findings on all pertinent radio-logical site suitability and environmental issues. See 10 C.F.R.
{ 50.10(e)(2). Subsequent proceedings will address the remaining portions of the application which are necessary for grant or denial of a Construction Permit.
The CRBRP was first authorized by the Congress in 1970 as a cooperative effort between industry and government to design, construct, and operate the nation's first demonstration-scale
*/  Citations to the record herein are in the following form:
a)    Applicants' Exhibit - A Exh; Staff's Exhibit -
S Exh; Intervenors' Exhibit - I Exh b)    Applicants' Witness - A W; Staff's Witness - S W:
Intervenors' Witness - I W c)    Transcript - TR d)    Citations to prefiled written testimony will include citations to exhibit number, page number, and transcript page.
s
 
LMFBR. Pub. L. No. 91-273, Section 106. In early 1972, the Atomic Energy Commission (AEC) accepted a joint proposal by the Commonwealth Edison Company (CE) of Chicago and the TVA to undertake the design, construction and operation of the demon-
                                                                                                    */
stration plant as part of the TVA electric system.-                                              Under this proposal, PMC, a non-profit corporation organized and existing under the laws of the District of Columbia, had the overall lead canagement responsibility for the CRBRP, TVA would operate it, and the AEC had lead technical responsibility for the nuclear reactor systems.          More than 750 electric systems in the United States have pledged more than $250 million in financial payments which are applied to the project by PMC.
In October of 1974, PMC and TVA jointly filed an appli-cation with the AEC for a Construction Permit and Operating License for the CRBRP pursuant to Section 104(b) of the Atomic
;      */  Pub. L. No. 92-84.                    See Joint Report of the House Committee
      ~~
on Science and Technology and the Joint Committee on Atomic Energy (JCAE), 94th Cong., 1st Sess. H. Rep. No. 92-294 at 32-35 (1975) [ hereinafter, Joint Report]; JCAE Authorization Report, 94th Cong., 1st Sess. S. Rep. No. 94-104 at 17-20 (1975) [ hereinafter, JCAE Report].
        **/  See Report on Hearings before the Joint Committee on Atomic Energy on the Basis for the Proposed Arrangement for the i            LMFBR Demonstration Plant, 92d Cong., 2d Sess. (Sept. 7, 8, and 12, 1972) [ hereinafter, JCAE Hearings] at IV-V.                                            See also Report on Hearings before the JCAE to Consider Proposed Changes in the Basis for the Cooperative Arrangement for Design, Construction, and Operation of the LMFBR Demonstra-tion Plant, 93d Cong., 1st Sess.                              (Feb. 28 and May 4, 1973).
        ***/ Id.
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                                                                                    ,_.._...-...-n    -      -
 
                                                          ^
i l
l i
Energy Act of 1954, a~s amended, 42 U.S.C. I 2011 et, seq. After the Energy Reorganization Act of 1974, 42 U.S.C. } 5801 et seq.,
transferred the developmental and regulatory functions of the AEC to the Energy Research and Development Administration (ERDA) and j  the NRC, respectively, the NRC assigned the application to its docket for review on April 11, 1975.
On June 18, 1975, receipt of the application and pro-caedings before the Atomic Safety and Licensing Board (the Board) were noticed. /        A timely joint petition for leave to intervene was filed by the Natural Resources Defense Council, Inc., the Sierra Club,  and the East Tennessee Energy Group- / (the Interve-nors), and on October 9, 1975, the petition was granted by this Board. The State of Tennessee Attorney General also filed a timely petition for leave to intervene and was admitted as a party on October 9,    1975. The City of Oak Ridge filed a petition for leave to intervene on July 17, 1975, amended that petition on 1
January 22, 1976, and was admitted as a party on March 4, 1976.
j */    40 Fed. Reg. 25708 (June 19, 1975).
l
  **/  The East Tennessee Energy Group is now defunct and the Natural Resources Defense Council, Inc. and the Sierra Club remain as joint intervenors.
  ***/ Roane County, which was admitted as a party ~by the Board's October 9, 1975 Order, was granted leave to withdraw from all participation by the Board's Order, dated December 13, 1976.
 
_. _ _ _ . _ _ - -            . _ _ _ .          . ._ _ - . _ =
On May 6,            1976, pursuant to authorization contained in the 1976 amendments to Pub. L. No. 91-273, as amended, the appli-cation was amended to include ERDA as a co-applicant (with PMC and TVA), and to reflect the realignment of the respective
]
project participants' roles. Under this realignment, ERDA assumed the lead management role in an integrated CRBRP project organiza-tion, which included PMC and TVA personnel, and TVA remained as the operator.          See Joint Report at 35; JCAE Report at 19; 122 Cong. Rec. S10613-22 (June 25, 1976); 122 Cong. Rec. H5835-5898 (June 15, 1976).              DOE is the successor-in-interest to ERDA. /
Commencing in November of 1975, extensive prehearing activities ensued,                  and by March of 1977 the NRC Staff had I            issued a Site Suitability Report (SSR) and Final Environmental Statement (FES).              S Exh 7.          On March 28, 1977, the Board issued an Order for commencement of LWA hearings in Oak Ridge on June
                */    See 42 U.S.C.            I 7101 et se~q.
                **/  Intervenors filed fifteen sets of interrogatories, seven sets of requests for admissions, and four requests for pro-duction of documents against the Applicants.                                                            Intervenors filed twenty-two sets of interrogatories, seven sets of requests for admissions, and three requests for production of documents against the RTC Staff. Appeals arose con-cerning the admissibility of two Intervenor contentions l                    which sought to litigate certain programmatic issues pre-viously considered in ERDA's LMFBR Program Environmental Statement (See United States Energy Research and Development Administration (Clinch River Breeder Reactor Plant), CLI                      13, 4 NRC 67 (1976)), and the untimely petition for leave to intervene of fourteen counties and municipalities in the vicinity of the site (See United States Energy Research and Development Administration (Clinch River Breeder Reactor l                    Plant), 4 NRC 383 (ALAB-383) (1976)).
l l
 
14, 1977, which hearings were to run continuously until comple. tion.
On April 20, 1977, the previous Administration announced its decision to cancel the project.                                        On April 22, 1977, i
,    ERDA filed a motion to suspend the proceedings, and on April 25,
;  1977 the Board issued an Order granting that motion.                                        In November of 1977, the NRC Staff suspended its review of the application.
In the ensuing four year period, the project continued design, research and development and procurement activities, while licensing activities remained suspended.                                          In each of those years, Congress acted to preserve the project by providing oubstantial funding. /
In August of 1981, the President signed the Omnibus Budget Reconciliation Act of 1981, Pub. L. No. 97-35, which expressed the intention that the project be expeditiously completed .-- /                  In a Nuclear Policy Statement of October 8,                      1981, the President directed that " government agencies proceed with a demonstration of breeder reactor technology, including completion of the Clinch River Breeder Reactor."                                          17 Weekly Comp. of Pres.
Doc. 1101-1102 (October 12, 1981).
    */              Pub. L. No. 95-240, March 7, 1978; Pub. L. No. 95-482, October 18, 1978; Pub. L. No. 96-86, October 12, 1979; Pub.
L. No. 96-369, October 1, 1980; Pub. L. No. 96-536, December 16, 1980; Pub. L. No. 97-12, June 5, 1981.
    **/              See H. Rep. No. 97-208, 97th Cong. 1st Sess. (1981);127 i
fong. Rec. S 8998 (1981); 127 Cong. Rec. H 5817-18 (1981).
    . - - , , . -      .e .            . _ _ _      ,_ -~-_m - - - -  --
 
On January 11, 1982, the Applicants filed a motion to lift the suspension of hearings, and on January 19, 1982, the Board granted this motion and issued a Notice of Prehearing Con-forence. On February 9-10, 1982, the Board held a prehearing conference, and February 11, 1982, issued an Order establishing a ochedule for all activities necessary for commencement of evidentiary hearings concerning LWA matters on August 23, 1982.
Pursuant to the Board's February 11, 1982 Order, all contentions related to the Construction Permit (CP) application were identified. The Intervenors restated and/or revised their original contentions, and filed additional contentions based upon new information. Upon consideration of the pleadings filed by the parties and two sets of prehearing meetings with the parties, the Board issued two Orders Following Conference with Parties, dated April 14 and April 22, 1982, which ruled upon the admissi-bility, scope, and applica 'lity (LWA vs. CP) of Intervenors' contentions, as follows:
a)    Contentions 4 (safeguards), 5 (site meteorology and population density and l                  risks to industrial facilities near the site), 6 (fuel cycle), 7 (alternative sites and designs), 8 (decommissioning), */ 11
  */    During a telephone conference between the Board and parties on December 7, 1982, Intervenors requested leave to withdraw contention 8, and the Board granted Intervenors' request.
 
1 b), c), and d) (genetic and somatic health ef fects of CRBRP operation and site suit-ability dose guidelines) were admitted for resolution in the LWA proceedings. Order Following Conference with Parties, April 14, 1982.
b)    Contentions 9 (emergency planning), 10 (sodium fires), and 11 a) (as low as reasonably achievable) were admitted, but Ceferred for the C.P. proceedings. */
Order Following Conference with Parties, April 14, 1982.
c)    Contentions 1, 2, and 3 (severe accidents in CRBRP) were admitted, subject to certain limitations on the scope of review for pur-poses of site suitability and environmental findings at the LWA stage. **/  Order
  --*/  Contention 1 (LWA procedure is inapplicable to first-of-a-kind reactors) was denied, as it presented an ultimate legal question for the Board following the taking of evidence.
Order Following Conference with Parties, April 14, 1982 at
: 3. Contentions 16 (radioactive river sediments), 17 (DOE planning for availability of fuel for CRBRP), and 22 (appli-cation of "as low as reasonably achievable" to accidents) were denied. Id. at 6-10. Contentions 20, 21, and 24 were withdrawn.      Id. at 9-10. Contention 18 (quality assurance) was denied wTthout prejudice to filing a contention with the i
requisite specificity and basis for the CP proceedings.        Id.
at 9.
  **/  On June 11, 1982, Intervenors' " Petition to Delineate the (Continued)
 
l Following Conference with Parties, April 22, 1982; S'ee Opinion, Intervenors' contentions 1,  2 and 3 (site suitability),
infra.
Extensive discovery ensued, during which all parties net the deadlines established by the Board's February 11, 1982
                                */
Prehearing Conference Order.-      On June 11, 1982, the NRC Staff issued its updated Site Suitability Report (SSR) which concluded that the Clinch River site was suitable for a reactor of the gen-eral size and type described in the application from.the stand-point of radiological health and safety (NUREG-0786). S Exh 1.
On July 13, 1982, the Advisory Committee on Reactor Safeguards (ACRS) issued a letter which supported the NRC Staff's site suit-ability conclusion. S Exh 4. On July 19, 1982, the Board issued a Notice of Evidentiary Hearing and Prehearing Conference, which ordered that hearings commence in Oak Ridge on August 23, 1982, and continue until completion of taking evidence on the issues and contentions admitted for the purpose of a limited work authorization hearing pursuant to 10 C.F.R.      50.10(e). On July Scope of the LWA Proceeding" sought direct review of this ruling from the Commission. The Petition was denied by the Commission by Order, dated November 17, 1982.
  */  By April 30, 1982, Applicants and Staff had updated their responses to Intervenors' 1975-77 discovery. As of the close of discovery on June 30, 1962, Intervenors had also filed four sets of interrogatories, four sets of requests for admissions, and three requests for production of docu-ments, and had deposed five persons from the NRC Staf f and eleven persons from the Applicants.
 
19, 1982, the NRC Staff issued and served upon all parties to the proceeding its update to the 1977 CRBRP FES (S Exh 7).        In issuing that document, the NRC Staff determined that it should be issued as a Draft Supplement to the 1977 FES, and that it should be recirculated for public comment before issuance as a Final Supplement.
As a result of the decision to recirculate the FES, the schedule for hearings ..antemplated by the Board's February 11, 1982 Prehearing Conference Order could not be met.      Upon consid-cration of motions filed by Applicants and Intervenors, and after hearing extensive argument during a conference with the parties, the Board issued an Order dated August 5, 1982 which scheduled hearings on radiological site suitability issues (portions of Intervenors' contentions 1,    2,  3, and Intervenors' contentions 2e)/ll d)1) and 2)), and ruled that hearings on environmental
                                                                */
issues would await issuance of the Final FES Supplement.-
After completion of site suitability hearings on August 23-27, 1982, the Board issued an Order establishing the schedule for completion of hearings on environmental issues. Pursuant      to this Order, the Board reopened discovery on all environmental I
  */    Intervenors filed with the Appeal Board a Petition for
  -~
Directed Certification in regard to this Order, which was denied on August 25, 1982. United States Department of Energy (Clinch River Breeder Reactor Plant), ALAB-688 (Memorandum and Order, August 25, 1982).
l
                                                            --m
 
t issuer / and set hearings for November 16-19, 1982, and December 13-17, 1982 to take evidence concerning the remaining environ-mental issues. Board Order, dated August ?l, 1982.
Neither the State of Tennessee Attorney General nor the City of Oak Ridge participated actively in the evidentiary hear-ings. By Order dated March 31, 1982, the Board granted the Attorney General's motion to withdraw as a party and participate as an interested State pursuant to 10 C.F.R. I 2.715(c), and by order dated September 7, 1982, granted the City's motion to withdraw as a party and participate as an interested municipality pursuant to 10 C.F.R. I 2.715(c)  The Board received the "Posi-tion Paper of the Tennessee Attorney General on Socio-Economic Impact Matters and other Matters Relating to the Clinch River Breeder Reactor Plant," dated November 10, 1982, and "The City of Oak Ridge's Statement Relative to the Socio-Economic Impact of the Clinch River Breeder Reactor Plant", dated November 12, 1982. At the direction of the Board (TR 3356-58; TR 7104), the Applicants and Staff filed Responses to the Attorney General's Position Paper and the City's Statement on January 11, 1983.
Neither the Attorney General nor the City conducted cross-examination, presented witnesses, or introduced documentary
    */    Pursuant to this Order, Intervenors filed one additional set of interrogatories against the NRC Staff (27th Set), and took eight depositions of more than twenty staff witnesses.
 
evidence concerning the socio-economic matters raised by their respective Position Paper and Statement.
On November 1,        1982 the NRC Staff issued the Final Supplement to the FES (FSFES).                        S Exh 8.        Evidentiary hearings on the subjects encompassed by Intervenors' environI intal conten-tions were held in Oak Ridge on November 16-19, 1982, and on December 13-16, 1982. Limited Appearance statements were received from 'uombers of the public in Oak Ridge during the hearing sest, ions held on August 23-27, 1982 and November 16-19, 1982.
Presentation of evidence on all LWA issues extended over the th. fee hearing sessions for a total of thirteen days, and was completed on December 16, 1982.- /
On December 16 and 17, 1982, and on January 4 and 5, 1983, the Board heard closing argument from all parties, speci-fically addressing the record evidence and disputed issues as to all LWA contentions.
The decisional record in this proceeding consists of:
a)    the Commission's Notice of Hearing, 1
  */    At least one Commissioner has aptly noted: "[t] hat the Licensing Board has managed to keep this proceeding focused and on track is almost a miracle and not something that could have been predicted by the Commission." United States Department of Energy (Clinch River Breeder Reactor                            Plant), -
CLI-83-1, (Memorandum and Order, January 6, 1983), Addi-tional View of Commissioner Roberts at 6. Applicants have attributed this less to miracles, which are outside the decisional record, than to the Board's management of the process.
 
b)  the material pleadings filed herein,
:                including the petitions and other plead-ings filed by the parties, and the Orders issued by the Board during the course of the proceeding; c)  the Exhibits received into evidence as indicated in Appendix A hereto; and d)  the transcript, consisting of 7105 pages (witnesses who testified are listed in Appendix B hereto).
In making its findings in this proceeding, the Board considered the entire record and all of the closing arguments and proposed findings submitted by the parties. Each of the proposed findings of the parties which is not incorporated directly or inferentially in this Partial Initial Decision is rejected as being unsupported in fact or in law or as being unnecessary to the rendering of this Partial Initial Decision.
This Board's jurisdiction is limited to a determination of findings of fact and conclusions of law on matters put into      ,
controversy by the parties to the proceeding or found by the 1
Board to involve a serious safety, environmental or common I
defense and security question. The Board has raised no such additional questions in the proceeding.
II. INTRODUCTION In connection with a limited work authorization under l  10 C.F.R. { 50.10(e), the Board is called upon to make findings
 
j                                  and determinations of two general types:    1) the findings required by 10 C.F.R.  $ 51.52(b) and (c) (environmental findings), and 2) a determination that, " based upon the available information and review to date, there is reasonable assurance that the proposed site is a suitable location for a reactor of the general size and type proposed from the standpoint of radiological health and safety considerations.    . . .        (site suitability determination). 10 C.F.R. { 50.10(e)(2).
The contested factual issues in the pt'ceeding can be aligned into nine sets or clusters of issues as follows:
A)  There are two sets of site suitability issues. These are encompassed by portions of Intervenors' contentions 1, 2,  and 3 (site suitability evaluation of severe accidents) and contentions 2e) and 11d)l) and 2) (site suitability dose guidelines).
B)  There are seven sets of environmental issues. These are encompassed by Inter-venors' contentions 2d), f), g) and h),
and 3c) and d) (environmental effects of severe accidents), 5b) (risk to nearby energy and national security facilities due to CRBRP accidents), 4 and 6b)4) l
 
(safeguards impacts), 6b)l) and b)3)
(fuel cycle impacts), Sa) and 7c)
(alternative sitos), 7a) and b) (likeli-hood of meeting programmatic objectives and alternative designs), and lib) and f
!                c) (genetic and somatic effects of CRBRP operation).
While there are specific matters of law which have a bearing on the atorementioned nine sets of contested factual issues, there is also one specific ultimate question of law sj presented for decision.-      Intervenors' original contention 1, which alleged that the LWA procedure is not applicable to first-of-a-kind reactors, presents an ultimate legal question to be determined in light of the evidence in the record as a whole.
See Order Following Conference With Parties, April 14, 1982 at 3.
III. OPINION In the Opinion which follows we proceed in sequence to address: A) the two sets of contested factual issues relating to
  */  The Position Paper of the Tennessee Attorney General and the Statement of the City of Oak Ridge present a question of fact and law concerning the need for additional socio-economic monitoring and license conditions for mitigation of impacts, based upon an uncontroverted evidentiary record in regard to socio-economic impacts of CRBRP construction.
Applicants and the NRC Staf f have previously filed Responses, dated January 11, 1983. Applicants rely on their January 11, 1983 Response for the purposes of this Proposed Decision and Findings of Fact.
 
l cite suitability, B) the seven sets of contested factual issues    f relating to environmental considerations, and C) the ultimate      !
question of law concerning the applicability of the LWA procedure.
A. SITE SUITABILITY ISSUES We address now the two sets of contested factual issues related to site suitability:      a) Intervenors' contentions 1, 2, and 3, which involve consideration of severe accidents in the site suitability evaluation; and b) Intervenors' contentions 2e) and lid)l) and d)2), which relate to the validity of the site cuitability dose guideline values.
: 1. Contentions 1,  2, and 3 (Site Suitability)
The parties are in agreerent that these contentions present three basic contested issues in regard to site cuitability matters:    a) whether hypothetical core disruptive accidents (HCDA's) should be considered as design basis accidents (DBA's) for CRBRP site suitability analysis; b) whether the site cuitability source term (SSST) for CRBRP results in radiological l
consequences which envelop the spectrum of DBA's; and c) whether
!  the containment design for CRBRP will reduce off-site doses to levels within the site suitability dose guideline values.
Finding 8.
As to the first issue, Intervenors have argued that HCDA's must be DBA's, while Applicants and the NRC Staf f have
 
arrived at a judgment that HCDA's should not be considered as DBA's for CRBRP site suitability analysis.                The Applicants and Staff base their position upon their analyses of the relevant initiators and sequences, and the general design characteristics, criteria, and state of technology for CRBRP, all of which show that CRBRP can be designed so that the likelihood that core conditions will progress to HCDA initiation is extremely low.
Findings 10-21.        Applicants' and Staff's evaluations of these factors are grounded upon deterministic engineering analyses and l
judgments, and rielther have relied upon quantitative probabilistic analyses at this juncture.              Idj Finding 40.
Intervenors have maintained that, for purposes of CRBRP cite suitability analysis, it is necessary to demonstrate through quantitative probabilistic analyses that the likelihood of an HCDA resulting in consequences exceeding the 10 C.F.R. Part 100 dose guideline values is less than 10-6 per reactor year.                As a l      corollary to this, Intervenors submit that this demonstration must be made on the basis of a complete, detailed safety review of the CRBRP.        Finding 40.
The disjunction in the parties' respective positions is primarily the product of their differing views concerning the proper scope of review at the LWA stage.              In our view, the site l      suitability determination at the LWA stage need not be a defini-tive, plant-specific or design-specific finding which requires a complete safety review for support.          The determination:
l 1
 
                                                                    )
: 1. does not require a complete safety review, but can be " based on the available information and review to date."-/
: 2. does not require definitive evidence, but only a showing of " reasonable assurance that the proposed site is a
                    ** /
suitable location.'
: 3. does not presuppose a completed, detailed design, but merely consideration of the general design characteristics of a " reactor of the general size and type proposed.'
The LWA decision is neither irrevocable nor with pre-judice to the succeeding safety review at the Construction Permit ctage. In this regard, the applicable NRC regulation states:
(4) Any activities undertaken pursuant to an authorization granted under this paragraph shall be entirely at the risk of the applicant and, except as to matters determined under paragraphs (e)(2) and (e)(3)(ii), the grant of the authorization shall have no bearing on the issuance of a construction permit with respect to the requirements of the Act, and rules, regulations, or orders promulgated pursuant thereto. ** * * /
Thus, should the subsequent safety review bring about a need for modifications in the facility or previous findirigs, the Applicants would bear the risk. This reinforces the notion that
*/    10 C.F.R. $ 50.10(e)(2)(iii).
**/    Id.
***/  Id. Comp'ar~e 10 C.F.R. $ 50.35(a), discussed below.
****/  10 C.F.R.  $ 50.10(e)(4).
 
information necessary for environmental and site suitability (LWA) findings can and should be substantially more limited than those for the CP, and that LWA findings can rest upon threshold considerations of design foasibility.
Not only is the LWA decision limited in scope, but even the subsequent CP review is subject to substantial limitations.
42 U.S.C. I 2235 ($ 185 of the Atomic Energy Act(AEA)) provides:
All applicants for licenses to construct or modify production or utilization facilities shall, if the application is otherwise acceptable to the Commission, be initially granted a construction permit.
Upon complet'on of the construction or modification of the facility, upon the filing of any additional information needed to bring the original application up to date, and upon finding that the facility authorized has been constructed and will operate in conformity with the application as amended and in conformity with the provisions of this Act and of the rules and regulations of the commission and in the absence of any good cause being shown to the Commission why the granting of a license would not be in accordance with the provisions of this Act, the Commission shall thereupon issue a license to the applicant. For all other purposes of this Act, a construction permit is deemed to be a ' license'.
The Supreme Court has directly addressed the scope of, and limitations upon, CP findings in Power Reactor ~Devel~opment Corp. v. International' Union' ~of El'ecurical', ' Radio ~ and' Mach'ine Workers (PRDC), 367 U.S. 396 (1961).        In PRDC the Court consid-ered the question of whether the Commission's safety finding at
 
                                                              "          information sufficient to provide the CP stage, i.e.,      . . .
reasonable assurance that a facility of the general type proposed can be constructed and operated at the proposed location without undue risk to the health and safety of the public," must be
! " backed up with as much conviction as to the safety of the final design of the specific reactor in operation as the second, final finding [i.e., for issuance of an operating license] must be."
PRDC, 367 U.S. at 407.        The Court concluded:
We think the great weight of the argument supports the position taken by PRDC and by the Commission, that Reg. 50.35 permits the commission to defer a definitive safety finding until operation is actually licensed. The words of the regulation themselves certainly lean strongly in that direction. The first finding is to be made, by definition, on the basis of incomplete information, and concerns only the " general type" of reactor proposed. The second finding is phrased unequivocally in terms of
            " reasonable assurance," while the first speaks more tentatively of "information sufficient to provide reasonable assurance."
The Commission, furthermore, had good reason to make this distinction. For nuclear reactors are fast-developina and fast-changing. What is up to daEe now may not, probably will not, be.as acceptable tomor-row. Problems which seem insuperable now may besolvedtomorrowgerhaptselfs process of construc ion            in the very Id.
The principles enunciated in PRDC, supra, remain valid today. The applicable NRC regulations define the scope of the CP                          l review as follows:
l l
 
I d
Sec. 50.35. Issuance of construction permits.
(a) When an applicant has not supplied initially all of the technical information required to complete the application and support the issuance of a construction permit which approves all proposed design features, the Commission may issue a construction permit if the Commission finds that (1) the applicant has described the proposed design of the facility, including, but not limited to, the principal architectural and engineering criteria for the design, and has identified the major features or components incorporated therein for the protection of the health and safety of the public; (2) such further technical or design information as may be required to complete the safety analysis, and which can reasonably be left for later consideration, will be supplied in the final safety analysis report; (3) safety features of components, if any, which require research and development have been described by the applicant and the applicant has identified, and there will be conducted, a research and development program reasonably designed to resolve any safety questions i            associated with such features or components; and that (4) on the basis of the foregoing, there is reasonable assurance that, (i) such                <
safety questions will be satisfactorily resolved at or before the latest date stated in the application for completion of con-struction of the proposed facility, and (ii) taking into consideration the site criteria contained in Part 100 of this chapter, the proposed facility can be constructed and operated at the proposed location without undue risk to the health and safety of the public. */
Thus, it is readily apparent that even the ultimate CP findings do not contemplate a final resolution of all safety issues. Rather, it is sufficient to find that certain issues can
    */
lo C.F.a. 5 so.3s(a).
 
i be left for later consideration, that research and development programs are reasonably designed to achieve timely resolution of  i those issues, and on this basis, there is reasonable assurance that, taking into consideration the site criteria contained in I
Part 100 of this chapter, the proposed facility can be constructed and operated at the proposed location without undue
  .      risk to the health and safety of the public. /
A fortiori, the Board's consideration of site i      cuitability issues at the LWA stage should be governed by the following principles:
i
: 1.      The analysis of site suitability should be based on (a) the available information and review to date, (b) a standard of reasonable assurance, and (c) a reactor of the general size and type proposed.
: 2.      The applicant proceeds at his own risk upon grant of an LWA, or even a CP.
: 3.      The review of safety issues should be undertaken at the CP stage, and even then, unresolved issues can await timely resolution at the OL stage.
      ~
        */    In contrast to 10 C.F.R. $ 50.10(e)(2), the CP finding under 10 C.F.R. { 50.35(a) contemplates a more specific analysis of  the facility, rather than findings concerning a reactor of  the general size and type proposed.
        **/  When carried to its logical extreme, Intervenors' position would hold that one cannot consider an LWA for CRBRP until the safety review is completed, thus rendering the LWA regulation a nullity for CRBRP. We addreds that issue separately in Section III C. below. More significantly
,            here, Intervenors' argument would suggest that one cannot I
even consider a CP for CRBRP.
1
: 4. LWA findings should be predicated upon feasibility of design measures, while detailed review of specific design measures
;          appropriate for the subsequent CP or OL stages.
With these principles in mind, the Board ruled that the ccope of inquiry at the LWA stage would be limited to whether it is feasible to design CRBRP to make HCDA's sufficiently improb-oble that they can be excluded from the envelope of design basis accidents for a reactor of the general size and type proposed.
Further, a full-scale inquiry into the specific design of the CRBRP is inappropriate at the LWA stage, and the review should be focused upon:  a) the major classes of accident initiators potentially leading to HCDA's; b) the relevant criteria to be imposed for the CRBRP; c) the state of technology as it relates to the applicable design characteristics or criteria; and d) the general characteristics of the CRBRP design (e.g.~, redundant, diverse shutdown systems). We also deferred consideration of Intervenors' contention Ib), which contemplated a detailed challenge to the Applicants' reliability program, and' Intervenors' contention 3a), which alleged that a detailed probabilistic risk assessment, comparable in scope to WASH-1400, was required. See Board Order of April 22, 1982.
On the basis of the evidence in the record, we now confirm that the foregoing limitations on the scope of inquiry at the LWA stage are appropriate, and find that the evidence is sufficient to show that HCDA's should not be DBA's for the
 
4 purpose of the CRBRP site. suitability determination. We base this finding on the weight of the reliable, probative evidence in the record, our interpretation of the proper role of probabilistic analysis at the LWA stage in light of this evidence, and our findings as to the opposing evidence advanced by Intervenors.
Findings 10-22; 27-41.
The CRBRP design has placed special emphasis on prevention of initiators and sequonces which could lead to HCDA conditions.      Finding 10. The record stands uncontroverted that Applicants and Staff have adequately identified the major classes of initiators potentially leading to HCDA's and, in addition, have identified the four major classes of design features which are necessary for prevention of the initiators and sequences that could lead to HCDA conditions.-l          Findings 12-13. The four classes of general design characteristics necessary for preven-tion of HCDA's are:      a) the reactor shutdown system; b) the shutdown heat removal system, c) features to prevent a double-cnded rupture of the primary heat transport system inlet piping,
    --*/  The Board is mindful of the Appeal Board's decision in Florida Power and Light Company (St. Lucie Nuclear Power Plant, Unit No. 2), ALAB-603, 12 NRC 30 (1980) but upon examination of the record, has found no evidence that there are important sequences which remain unidentified, and which would suggest a need for additional inquiry and scrutiny.
Moreover, Intervenors have not suggested any important sequences of this nature, or any which might fall within the I          ambit of the Commission's TMIul decision. Metropolitan Edison Company (Three Mile Island Nuclear Station Unit No.
1), CLI-80-16, 11 NRC 674 (1980).
 
4 and d) features and capabilities to limit local imbalances between heat generation and heat removal in the core.
Finding 14.
The general design characteristics, criteria, and state of technology as to each of these systems and features are such that it is clearly feasible to design CRBRP so that the likeli-hood of progression to HCDA conditions is extremely low.      The reactor shutdown system will consist of two fast-acting, redundant, independent, and diverse systems (in contrast to one such system in current generation light water reactors (LWR's)),
which are based upon proven technology and are each capable'of independently shutting down the reactor. The shutdown heat removal systems are likewise designed with redundant, indepen-dent, and diverse features, and provide the capability for removing decay heat through any of four separate paths using components which are based upon well-established technology.      The CRBRP will operate at a low pressure (the system is operated nominally at atmospheric pressure; pressure increases only as a result of operation of the primary heat transport system pumps),
which is well below the saturation pressure of the sodium coolant at operating temperature. In spite of this, a rupture at a l
cection of the primary heat transport system piping near the reactor vessel inlet could result in a flow reversal in the core, reduced heat removal, and could potentially lead to initiation of l
HCDA conditions. Controls, analyses, and technology programs
 
have been implemented to assure that piping material properties will not degrade in-service, and that the likelihood of an undetected leak in this section of the inlet piping in excess of design basis values is extremely low.            To assure that local imbalances between heat generation and heat removal will not spread and involve a significant portion of the core, the design includes a multiplicity of redundant core inlet flow paths and sensitive detection systems, and incorporates results of substantial experience and analyses from EBR-II and worldwide LMFBR development, which show that the likelihood of fuel failure propagation from local areas to core-wide involvement is extremely remote. /            Findings 10-18.
We find that the approach taken by Applicants and Staff to the engineering and analysis of these systems and features, coupled with their use of sound deterministic criteria and judg-ment, provides a high degree of assurance that the likelihood of progression to HCDA conditions can be made extremely low.
Findings 10-21.
  */    The mechanical design of the inlet flow paths provides redundant flow paths in the fuel sub-assembly inlets, in the inlet modules that hold groups of sub-assemblies, and in the core support structure that holds and supports the inlet modules. This design will effectively preclude the recur-rence of events similar to that at the Fermi LMFBR, where a loose plate in the primary system blocked flow to fuel assemblies and resulted in partial melting of two sub-assemblies. Findings 18 and 31.
4
 
l i
Our confidence in this finding is bolstered by the fact that in spite of the emphasis Which has been placed on the prevention of HCDA conditions, both Applicants and Staff have considered, and the design incorporates, specific provisions for mitigating the consequences of events beyond the design basis, i.e., HCDA's. The design will include an annulus cooling system, a containment vent / purge and clean-up system, and associated instrumentation to relieve containment pressure and limit hydrogen concentrations. In addition, the reactor coolant boundary has been strengthened and seals have been added to the reactor vessel closure head to accommodate loadings from core disruption and limit the leakage of sodium gases and vapors to the environment. These provisions have been evaluated against a broad range of core disruptive and core-melt accident conditions, including consideration of all pertinent nuclear, thermal, structural and radiological factors, to ensure that the residual risks of such accidents in CRBRP can be made acceptably low and comparable to those in LWR's.      Finding 22.
Our finding is also supported by the fact that Appli-cants have proposed, and the Staff will impose, a confirmatory reliability program Which will assure that the potential for reliability inherent in the redundant, diverse, and independent design features is, in fact, realized.      Findings 10-19. On the basis of the record, we find that this is the appropriate and logical role for the reliability program.      Findings 10-19;
 
r I
Finding 40.          Intervenors' position, of course, is in fundamental disagreement.          In the Intervenors' view, the proper role of the
:  reliability program should consist of an 31 priori demonstration that the design and plant hardware will, on the basis of probabilistic analyses, satisfy a single-valued numerical reliability objective.            Finding 40.
The Board finds that a decision of this moment should rest upon the reliable, probative evidence of the deterministic sugineering considerations advanced by Applicants and the NRC Staff.          The reliability program has an appropriate role in terms of assuring that the reliability inherent in the design is realized, and detailed inquiry concerning the adequacy of this program is obviously open to consideration in subsequent proceedings.          At the same time, however, the Board finds that, consistent with Commission policy and practice, the state-ot-the-art in regard to quantitative probabilistic analysis is not sufficiently advanced to provide a reliable basis for this decision at this time. /            All parties agree that the decision as
                                                  .              e
                                                        <        J
  -*/          The Commission's proposed policy statement on safety goals for nuclear power plants states:                    -
Since the completion of the reactor safety study, further progress in developing probabilistic risk assessment and in accumulating relevant data has led to the recognition that it is feasible to begin to use quantitative reactor safety guidelines for limited purposes. .However, because of the sizable uncertainty still present in the methods and gaps in the data base --
(Continued)                                '
 
to whether an HCDA must be a DBA must be founded on judgment (Finding 40), and we find that the judgments of the Applicants and Staff are supported by sound engineering concepts and the weight of the evidence in the record.      Findings 10-22, 40.
We turn now to consideration of Intervenors' opposing.
ovidence. Having disposed of Intervenors' primary argument concerning the role of reliability analysis and probabilistic risk assessment in this decision, Intervenors advance three additional lines of historical argument involving previous domestic sodium cooled reactor, foreign sodium cooled reactor, and CRBRP experience to urge that an HCDA should be a DBA.
essential elements needed to gauge whether the guidelines have been achieved -- the quantitative guidelines should be viewed as aiming points or numerical benchmarks which are subject to revision as further improve-ments are made in probabilistic risk assessment. In particular, because of the present limitations and the state of the art of quantitatively estimating risk, the numerical guidelines are not substitutes for existing regulations.
47 Fed. Reg. 7023, 7024 (Feb. 17, 1982)
Further, in considering the role of probabilistic risk assessment in Construction Permit applications for light water reactors, the Commission has not required the per-formance of a probabilistic risk assessment as a condition to issuance of a CP, but instead has required completion within two years after CP issuance. 47 Fed. Reg. 2286, 2302 (Jan. 15, 1982). We note here that, consistent with the foregoing, a detailed probabilistic risk assessment is being performed for CRBRP and that completion is scheduled for 1984. S W Morris, TR 1984, 5649-50. As with the reliability program, this analysis will serve as a valuable tool to augment the deterministic analyses and engineering judgments which form the basis for decision at this juncture.
 
                                                                      ~
\
Intervenors rely upon domestic reactor experience that is outdated, inapplicable, or already anticipated in the CRBRP design. Findings 28-34. The foreign reactor experience cited by Intervenors, on the basis of the record as a whole, is consistent with the approach proposed, by the Applicants and NRC Staf f.
Findings 35-37.
Intervenors point out that the 1975 CRBRP parallel design treated an HCDA as a DBA. Intervenors extend this fact to argue that the parallel design treated an HCDA as a DBA at the insistence of the NRC Staff. The record shows that in.1974 the NRC Staf f adopted a "show-me" attitude vis-a-vis the Applicants' position that an HCDA should not be a DBA. Yet Intervenors characterize the Staff'a {{letter dated|date=May 6, 1976|text=May 6, 1976 letter}}, which indicated that HCDA's can and should be excluded as DBA's for CRBRP, as a
  " dramatic reversal" of position. The record shows, however, that all Staff statements expressing skepticism as to whether Applicants could make a convincing case were made prior to the time that the application was docketed for review. The {{letter dated|date=May 6, 1976|text=May 6, 1976 letter}} did not signal a dramatic reversal of position, but was merely the Staff's first educated judgment on this issue after having had an opportunity to review the design. Finding
: 39. For all of these reasons, the Board rejects Intervenors' historical arguments. Findings 28-39.
On the basis of the foregoing, the Board finds that the deterministic approaah adopted by the Staff and Applicants is
 
both reasonable and appropriate, that the role urged by Intervenors for reliability analysis and probabilistic risk assessment is inappropriate and in conflict with Commission practice and policy, and that Intervenors' historical lines of argument are without merit.              Accordingly, the Board finds, on the basis of the reliable, probative evidence in the record as a whole, that the Applicants have sustained their burden of showing that HCDA's need not be DBA's for the purposes of CRBRP site cuitability analysis.
Given our decision concerning the definition of i        the design basis envelope for CRBRP site suitability analysis,                                we find essentially no dispute in the record concerning the second
!      contested issue -- whether the site suitability source term (SSST) release recommended by the NRC Staf f for purposes of site suitability analysis results in hazards not exceeded by any accident considered credible.                      See 10 C.F.R. { 100.ll(a)(2),
n.l. The evidence of record is uncontroverted that the SSST represents an assumed release from the core whose consequences envelop and are not exceeded by any accident considered credible, i.e.,  any release from the core within the design basis envelope. Finding 23.
It remains to consider the third contested issue --
whether the containment will limit doses from the SSST release I        within the site suitability dose guideline values.                            In this
 
l regard, Intervenors have advanced four primary arguments con-testing the validity of the methods of analysis and assumptions used by Applicants and the NRC Staff for site suitability dose calculations.-/ These are: a) the dose calculations should consider the effects of the " entire passage of the cloud"; b) the dose calculations should have assumed the use of a plutonium source term with higher relative concentrations of the plutonium isotopes Pu-238 and Pu-241; c) the dose calculations did not employ appropriate dosimetric models; and d) the dose calcula-tions did not include consideration of the vent / purge and clean-up system provided in the design for beyond-design basis events.
10 C.F.R. } 100.ll(a)(2) contemplates thtt the dose at the low population zone boundary will be calculated for "tha entire passage of the radioactive cloud." 10 C.F.R. $ 100.ll(a)(2) The NRC Staff's calculations did not extend beyond a 30-day period, based on their conclusion that the dose contribution after a 30-day period would be negligible.            Intervenors assert that an NRC Staff sensitivity calculation, which assumed an instantaneous release of the site suitability source term inventory remaining in the containment at the end of the 30-day period, showed doses
  */
Intervenors have also argued that an HCDA should be a DBA and thus the site suitability source term and resultant dose consequences should have higher values in order to envelop the consequences of an HCDA. Having disposed of this argu-ment previously, we need not return to consider it in the context of the methods of analysis and assumptions used by the Applicants and NRC Staff for site suitability analysis.
 
l
[ cignificantly larger than those calculated for the first 30 days.
This sensitivity calculation, however, was an attempt by the Staff to define the upper bound of effects attributable to releases beyond the first 30 days. The Applicants' analyses showed that ninety percent of the total 30-day dose would be incurred in the first day and ninety-eight percent would be incurred in the first week. Moreover, the Staff's sensitivity calculation did not consider the effects of plateout and fallout within the containment after the first 24 hours. When the Staff recalculated the dose effects after 30 days, assuming a reason-able level of plateout and fallout and a design basis leak rate consistent with the Staff's normal site suitability source term analysis assumptions, the doses calculated were reduced to an insignificant fraction of the dose calculated for the 30-day period. The doses from the NRC Staff's initial sensitivity l
analysis, which are nevertheless within the site suitability dose guideline values, are an extremely conservative upper bound analysis, and do not invalidate the Staff's conclusion that the dose contribution beyond 30 days would be negligible. Finding 43.
Intervenors argued that the NRC Staff underestimated the relative concentrations of the plutonium isotopes Pu-238 and Pu-241 for the site suitability source term, since the future use of high burn-up LWR fuel or repeated recycle of fuel in CRBRP would result in an increase or build up of the isotopes Pu-238
 
and Pu-241 relative to the values assumed in the Staff's analysis.                            Since the radioisotopes Pu-238 and Pu-241 are more radiotoxic than the predominant plutonium isotope in the CRBRP fuel (Pu-239), Intervenors argue that the doses calculated by the Staff and Applicants are non-conservative.                            Intervenors' argument is without merit for four reasons.                            First, the values used by the NRC Staff for SSST analysis assumed Pu-238 and Pu-241 concentra-tions significantly greater than those contemplated by the appli-cation.                    Second, there is ample low-burn-up LWR fuel available which falls within the limits of the isotopic concentrations assumed in the Staff's analysis.                            Third, in the event that the Applicants consider the use of fuel with higher concentrations of Pu-238 and Pu-241 in the future, such that the limits of the site suitability source term analysis might be exceeded, the matter would be reviewed by the NRC Staff and a license amendment undertakea as necessary and appropriate.                          Fourth and finally, Intervenors' argument misconceives the nature of the physics characteristics of CRBRP.                          Unlike an LWR, where the thermal neutron spectrum and repeated recycle would result in build-up of the isotopes Pu-238 and Pu-241, the fast neutron spectrum in CRBRP would cause these isotopes to be burned or reduced in relative concentration upon repeated recycling.                            Thus, the Staff's assumed values for plutonium isotopic concentrations in the site suitability source term analysis are appropriately conservative, susceptible to licensing review and amendment if
 
35 -
and as necessary, and not affected by the future prospect of recycling fuel in the CRBRP.      Finding 44.
Intervenors argued that the models used by the NRC Staff for SSST analysis did not use the most advanced dosimetric models for calculating SSST doses, and to that extent, the analysis underestimated bone doses by a factor ranging between two and three. The models used are those which are normally applied for this purpose in the NRC regulatory process.      The models have their origins in ICRP-2, and ICRP-2, in turn, provides the basis for the Commission's existing radiation protection standards in 10 C.F.R. Part 20. Most importantly, additional calculations performed by both Applicants and Staff using the more recent models urged by Intervenors showed doses well within the site suitability dose guideline values.      Finding 45.
Intervenors also argued that the Staff's SSST analyses underestimated the doses by failing to concider releases through the vent / purge and clean-up system provided in the design for beyond-design basis events.      In the simplest form of expression, Intervenors argue that if one considers the system that pumps radioactivity back into the containment (the annulus exhaust /
filtration system provided as an engineered safety feature for design basis events), one should also consider the system that pumps radioactivity out of the reactor containment building (the vent / purge system for beyond-design basis events). The site suitability source term analysis was properly predicated on the
 
assumption that an HCDA should not be a DBA. The site suit-Ebility source term analysis assumed that the containment annulus cxhaust/ filtration system would be available as an engineered cafety feature for mitigating design basis events. After containment isolation, the annulus exhaust / filtration system would take any leakage from the reactor containment building, pass that leakage through a filtration system, and recirculate that leakage into the containment annulus between the reactor containment building and the confinement building. It does not recirculate that leakage back into the reactor containment building as presumed by Intervenors. On the other hand, if conditions should progress beyond the design basis, and containment integrity were threatened during a time period of about a day after initiation of the event, the operator could then manually open a normally closed containment vent (the vent / purge system to which Intervenors refer) which would dis-charge to the environment through a containment clean-up system. Under design basis conditions, the containmant vent / purge system is closed and has no meaningful physical role. Therefore, it is neither meaningful nor necessary to consider the effects of the vent / purge system in the context of the CRBRP SSST analysis. Finding 46.
Upon conc.deration of Intervenors' four arguments, the Board finds that nene invalidates the SSST analyses performed by the NRC Staff and Applicants. Findings 42-46. The evidence in
 
the record clearly shows that the CRBRP containment will limit consequences within the dose guideline values selected for CRBRP site suitability analysis.      Findings 24-25; Findings 68-69.
Accordingly, in regard to the three contested site cuitability issues encompassed by Intervenors' contentions 1,      2, and 3,  the Board finds that a) an HCDA need not be a DBA for purposes of the site suitability analysis, b) that the conse-quences associated with the site suitability source term release celected by the NRC Staff for CRBRP site suitability analysis envelop the consequences of any design basis accident, c) that the Staff's and Applicants' site suitability analyses were adequate, and d) the containment will limit site suitability cource term doses within the site suitability dose guideline l
values.      Findings 7-46; Findings 68-69.
: 2. Contention 2e) (Site Suitability Dose Guidelines)
Intervenors' contention 2e) alleges that the site suitability dose guideline values recommended by the NRC Staff for CRBRP site suitability analysis are not valid.      The existing NRC regulations specify site suitability dose guideline values of 300 rem for thyroid exposure and 25 rem for whole body exposure.
f  10 C.F.R. I 100.ll(a), n.2.      The NRC Staff developed supplemental guideline values for organs of importance to plutonium exposure for the purposes of CRBRP site suitability analysis. 10 C.F.R.
  $ 100.11(a), n.2, expresses the purpose of the site suitability dose guidelines as follows:
 
                                              . .  . Neither its [the whole body dose of 25 rem] use nor that of the 300 rem value for thyroid exposure as set forth in the site criteria guidelines are intended to imply that these numbers constitute acceptable limits for emergency doses to the public under accident conditions. Rather, this 25 rem whole body value and the 300 rem thyroid value have been set forth in these guides as reference values, which can be used in the evaluation of reactor sites with respect to potential reactor accidents of extremely low probability of occurrence, and low risk of public exposure to radiation.
In order to derive supplemental guideline values for CRBRP site suitability analysis, the NRC Staff took the existing i
dose guideline values in 10 C.F.R. Part 100 as given, and applied mortality risk weighting factors from ICRP Publication Number 26 (ICRP-26 ) to scale from the existing 10 C.F.R. Part 100 whole body and thyroid values, and thus derive two sets of supplemental guideline values for the other organs. Finding 48.
1 The Staff selected the lowest (i.e., the most conser-vative) set of values, which were the values derived from scaling with ICRP-26 weighting factors from the existing 10 C.F.R. Part 100 300 rem thyroid value. The itaff then reduced these values by a f actor of two for the purpost s of the construction permit review in order to account for uncertainty. The values derived by the NRC Staf f for purposes of CRBRP site suitability analysis are well supported by the available body of scientific evidence, the existing 10 C.F.R. Part 100 dose guideline values, and the stated purpose of those dose guideline values.      Findings 47-49.
 
Intervenors have argued that more conservative dose guideline values should have been adopted since:                                            a) one should apply the non-stochastic limit of 50 rem per year set forth in ICRP-26, as well as the mortality risk weighting factors set forth in ICRP-26, b) one should derive the doses by applying weighting factors based upon the EPA environmental radiation protection requirements for normal operation of activities in the uranium fuel cycle, and c) the dose guideline values should be reduced by a factor greater than two to account for uncertainties in the dose and health effects models.                                            Finding 50.
ICRP-26 establishes a non-stochastic limit of 50 rem per year for occupational exposures to any organ.                                              Non-stochastic offects are those health effects which show a threshold effect with exposure to ionizing radiation (e.g., cataracts), while stochastic ef fects are those which occur in an apparently random nanner and do not show a threshold with exposure to radiation (e.g., cancer).                                          Application of the non-stochastic limit to the derivation of CRBRP site suitability dose guidelines values is not appropriate for two reasons.                                            First, this limit corresponds to an annual occupational dose, and its direct use in deriving dose guidelines would produce values which are higher than those set forth in 10 C.F.R. Part 100 or those recommended by the NRC Staff. A dose of 50 rem per year, applied over a 30-year operating lifetime at CRBRP, would yield thyroid and lung values of 1500 rem.            In contrast, the Staff's recommended values include
 
i
,    the existing 300 rem thyroid value in 10 C.F.R. Part 100, and a I
value for lung of 75 rem.      Secondly, even if the 50 rem per year non-stochastic limit were artificially limited to a one-time exposure, application of the 50 rem per year limit would require reducing all dose guideline values (except whole body), including the '10' C'.F.R'. ' Fart'100 300' rem' thyroi~d' value, to 50 rem. This, of course, would challenge the validity of the existing regula-
                                      */
tion in 10 C.F.R. Part 100 .--      Therefore, the Board finds that application of the non-stochastic limit from ICRP-26 would not be appropriate for use in deriving dose guideline values for CRBRP site suitability analysis.      Findings 51-53.
The EPA environmental radiation protection requirements for normal operation of activities in the uranium fuel cycle specify a value of 25 mrem per year for the whole body and 25 mrem per year for any other organ for exposure to the public.
      */
10 C.F.R. $ 2.758(a) forecloses attack on Commission rules or regulations in individual licensing proceedings. See, e.g., Metropolitan Edison Company (Three Mile Island Nuclear Station, Unit No. 2), ALAB-456, 7 NRC 63 (1978) (challenge to fuel cycle rule); Cleveland Electric Illuminating Company (Perry Nuclear Power Plant, Units 1 and 2), LBP-81-24, 14 NRC 175 (1981) (challenge to regulation on reactor vessel integrity); Sacramento Municipal Utility District (Rancho Seco Nuclear Generating Station), LBP-81-12, 13 NRC 557 (1981) (challenge to standards for operator training);
Pennsylvania Power & Light Company (Susquehanna Steam Electric Station, Units 1 and 2), LBP-79-6, 9 NRC 291 (1979)
(challenge to reliance on " single failure" criterion of Part 50, App. A); Philadelphia Electric Company (Peach Bottom Atomic Power Station, Units 2 and 3), LBP-75-22, 1 NRC 451 (1975) (challenge to emergency core cooling system criteria).
 
Using this analogy, Intervenors argue that site suitability dose guideline values of 25 rem for the whole body and 25 rem for every other organ should be applied in the CRBRP site suitability cvaluation.          Finding 54.
;                        Application of the EPA requirements in deriving the i
cite suitability dose guideline values would yield a value of 25 ren for thyroid, thus challenging the existing 10 C.F.R. Part 100 300 rem dose guideline value.                    Further, the use of the EPA requirements as an analogy would ignore the best available scientific evidence, which shows that the ICRP-26 mortality risk weighting factors describe the relative radiosensitivities of' the various human organs in an appropriate fashion.- Finally, the bases for derivation of the EPA requirements are inconsistent with their application in deriving 10 C.F.R. Part 100 guideline values.      The EPA requirements are based on a cost / benefit balance, and there is no evidence in the record to show how that balance was struck, or how it reflects the best available scien-tific evidence.            In addition, the EPA requirements are intended to " encompass abnormal but anticipated releases of radioactive caterial to the environment associated with effluent control measures, [but] potential releases associated with the pos-sibility of accidents involving the nuclear safety of the facilities are beyond the scope of the proposed rulemaking, which is limited to environmental radiation due to normal operation."
l 39 Fed. Reg. 16906 (May 10, 1974) (emphasis added).                                          Thus, the
 
i intent of the EPA regulations is entirely inconsistent with their application in deriving dose guideline values for CRBRP site i          suitability evaluation, and with the stated purpose of the 10 C.F.R. Part 100 dose guidelines.                                          Findings 54-56.
Intervenors argue that the Staff's reduction of the
;          dose guideline values by a factor of two to account for uncer-I tainties at the Construction Permit stage is non-conservative in
:          light of three major sources of uncertainty in the dose and health effects models:                                      a) the " hot particle" hypothesis, b) the
,          " Morgan bone" hypothesis, and c) the " warm particle" hypothesis.
Finding 57.
l                                        The " hot particle" hypothesis has been considered and rejected by the overwhelming consensus of scientific opinion.                                                      It
                                                                                                                        */
has also been rejected by the Commission and the EPA.'-
Intervenors have even stipulated that there is not much support in the written literature for the hypothesis beyond that expressed by its authors.                                        Finding 58.
The " Morgan bone" hypothesis argues that the value of I          maximum permissible body burden for Pu-239 set forth in ICRP l
l Publication Number 2 (ICRP-2) is non-conservative by a factor of 240.            The dose guideline values recommended by the NRC Staff for organs of importance to plutonium exposure were not derived on the basis of ICRP-2, but only on the basis of the existing 10
          */            41 Fed. Reg. 15371 (April 12, 1976); 42 Fed. Reg. 1288 (January 6,                      1977).
_ . . -  _ . _ _ _ _ _  , _ _ _ - - - _      . _ . . _ _ .  - - _ -_          ~ _ _ _ _ _ _ _ _ _ . . _ , _ . -        .-
 
C.F.R Part 100 dose gaideline values and the ICRP-26 mortality risk weighting factors.                      Although the record nevertheless shows that the Morgan hypothesis should not have a substantial effect on the validity of the ICRP-2 maximum permissible body burden value (a factor of two rather than 240), it does not affect the validity of the Staff's recommended dose guideline values since their derivation was independent of ICRP-2.                          Finding 59.
i                    Although the record is barren of any evidence to suggest that there is a logical nexus between the " warm particle" hypothesis and the validity of the dose guideline values, the record nevertheless shows that the " warm particle" hypothesis is speculative and unsupported by the available scientific eridence.
Finding 60.              Accordingly, the record shows adequate consideration of all uncertainties.                  The Staff's factor of two reduction in the dose guideline values to account for uncertainties is prudent, and the dose guideline values recommended by the NRC Staff are i
l appropriate for purposes of CRBRP site suitability analysis.
l Findings 57-60.
B.            ENVIRONMENTAIi ~ ISSUES There are seven sets of contested factual issues in regard to environmental matters within the scope of NEPA. These are Intervenors' contentions:                          a) 2d), f), g), and h), and 3c) and d) (environmental effects of severe accidents); b) 4 and 6 b)4) (safeguards impacts); c) Sb) (risk to nearby energy and national security facilities due to CRBRP accidents); d) 6b)l)
 
i i
and b)3) (fuel cycle impacts); e) Sa) and 7c) (alternative cites); f) 7a) and b) ( likelihood of meeting programmatic objectives and alternative designs); and g) lib) and c) (genetic and somatic effects of CRBRP operation). We turn now to address each of these issues.
: 1. Contentions 2d), 2f), 2g), 2h), 3c) and 3d)
(Environnantal Effects of Severe Accidents)
The broad issue raised by these Intervenors' conten-tions concerns whether or not the NRC Staff has adequately assessed the environmental impacts of accidents in the Final Environmental Statement (FES) and Final Supplement to the Final Environmental Statement (FSFES). S Exh's 7 and 8. There is no dispute in the record concerning the adequacy of the Staff's analysis of the environmental effects of design basis accidents for CRBRP. Finding 75. Rather , the disputes in the record rolate to the adequacy of 5.he Staff's analysis in Appendix J of the Final Supplement, which concluded that the risks of severe, beyond-design basis accidents in CRBRP are acceptably low and comparable to those of LWR's.
The Staff's Appendix J analysis was intended to satisfy the requirements of the Commission's Statement of Interim Policy concerning Nuclear Power Plant Accident Considerations under the National Environmental Policy Act of 1969.      45 Fed. Reg. 40101 (June 13, 1980); S Exh 8 at J-1. The Commission's Statement of Interim Policy provides that Environmental Impact Statements
 
shall include consideration of the site-specific environmental impacts attributable to accident sequences that lead to releases of radiation and/or radioactive materials, including sequences that can result in inadequate cooling of reactor fuel and to neiting of the reactor core.          In this regard, attention shall be given both to the probability of occurrence of such releases and to the environmental consequences of such releases.              45 Fed. Reg.
40101.          The Statement of Interim Policy further contemplates that: a) the releases considered will include radiation and/or radioactive materials entering environmental exposure pathways, including air, water and ground water; b) events or accident i cequences that lead to releases shall include, but not be limited to, those that can reasonably be expected to occur; c) in-plant accident sequences that can lead to a spectrum of releases shall be discussed and shall include sequences that can result in I
inadequate cooling of reactor fuel and melting of the reactor I
core; d) to the extent that events arising from causes external l
to the plant are considered possible contributors to the risk associated with the particular plant, they shall also be dis-cussed; e) detailed quantitative considerations that form the basis of probabilistic estimates of releases need not be incorporated in Environmental Impact Statements, but shall be referenced therein; f) the environmental consequences of releases whose probability of occurrence has been estimated shall also be discussed in probabilistic terms (such consequences shall be
 
characterized in terms of potential radiological exposures to individuals, to population groups, and, where applicable, to biota); g) health and safety risks that may be associated with oxposures to people shall be discussed in a manner that fairly reflects the current state of knowledge regarding such risks; h) cocio-economic itpacts that might be associated with emergency measures during or following an accident should also be discussed; i) the environmental risk of accidents should also be compared to and contrasted with radiological risks associated with normal and anticipated operational releases; and j) major uncertainties in the probabilistic estimates shall be discussed. Id. at 40102.
Upon comparison of Appendix J to the foregoing elements of the Commission's Statement of Interim Policy, the Board finds that the scope, content, and analysis of Appendix J satisfy each of those elements. Finding 76.      Moreover, the Staff's Appendix J analysis conservatively estimates ( i~. e~. , o v e r e s t i m a t e s ) the risks 1 of core melt and disruptive accidents in CRBRP.                    Findings 72-87.
Intervenors have specifically argued that the Staff's Appendix J analysis is inadequate because: a) it has underesti-mated the frequency of core degradation due to a loss of heat sink (LOHS) event; b) it has underestimated the frequency of pipe rupture and erroneously concluded that pipe rupture is not a dominant contributor to the estimated LOHS frequency and c) it l
1
 
has underestimated the frequency of containment failure.
Finding 81.
The NRC Staff's Appendix J analysis assigned a fre-
                      -4 quency of 10        per year for core degradation due to LONS. The Staff estimated this value by reference to pressurized water reactor (PWR) reliability experience, Which indicates that auxil-iary feedwater (AFW) system failures dominate the frequency of LOHS events.      Although the CRBRP has a back-up Decay Heat Removal Service (DHRS) Which does not depend upon AFW, so that the decay heat removal function in CRBRP should be at least, if not more, reliable than that of a PWR, the Staff nevertheless estimated the CRBRP AFW    system failure frequency at the high side of the LWR range. In artificially overestimating the frequency of AFW cystem failures, the Staff made other potential contributors to LOHS frequency, such as fuel failure propagation or pipe rupture, at most small fractions of the dominant contributor (AFW) to the estimated LOHS frequency.        Finding 82.
Intervenors argued that the LWR AFW reliability studies chow higher failure frequencies than those estimated by the NRC Staf f in Appendix J.      Intervenors cited the Calvert Clif fs AFW reliability study as one example of this, but on cross-examination it was established that this was the only example of which Intervenors were aware.      In fact, the record clearly shows that the CRBRP AFW system is very likely to be substantially more reliable
                                                              -4 than the Calvert Cliffs system, and that the 10          failure
 
l frequency estimated by the NRC Staff resides at the high end of the range one can associate with LWR experience.                            Finding 83.
Based upon PWR experience and the design characteris-tics of CRBRP, the NRC Staff concluded that the frequency of primary pipe rupture would be at most a small fraction of its
                                            -4 conservatively estimated 10                  per year LOHS frequency.            In reliance upon a report by a Dr.. Harris, Intervenors argued that n
the frequency of primary pipe rupture ~may be twelve times higher than that for a PWR. There are three infirmities in Intervenors' argument. First, Intervenors' witness Dr. Cochran admitted on cross-examination that he is not an expert in regard to pipe rupture probability. Second, the report by Dr. Harris, upon which Intervenors relied, incorporates the explicit conclusion that the probability of pipe rupture for CRBRP is 0.1--l times                          .
that for a PWR.            Third, subsequent work by Dr. Harris shows that the absolute probability of a pipe rupture in a PWR is in the
                      -7            -8 range of 10              to 10      per year.        By necessary inference from Dr. Harris' two articles, the probability of pipe rupture in
                                          -7            -9 CRBRP would range from 10                  to 10            per year, and would
                                                                                            -4 constitute a very small fractional contributor to the Staff's 10 estimate for LOHS frequency.                Finding 86.
Finally, Intervenors argued that the loss of on-site and off-site power could cause a breach of containment through a loss of containment mitigating systems (principally the beyond-design basis annulus cooling and vent / purge systems),                      and that i    --    - - . _ .    -      ,_        ,            _ _ _
 
the Staff had not accounted for this failure mode in estimating the frequency of containment failure. The record shows, however, that ths St'aff's Appendix J analysis did not assume that any of these mitigating systems would be available for about a day after initiation of the event, and thus its conclusions as to contain-ment failure frequencies would not be affected by consideration of this failure mode. In addition, the record shows that the Staff did consider all relevant contributors to the frequency of containment failure by overpressure from loss of the mitigating systems, that it conservatively downgraded the capability of these systems, and that it overestimated the frequency of failure by overpressure. In spite of this, Intervenors argued that an article in Nuclear ~~ Safety showed that the frequency of breach of containment should be higher by a fdctor of 10 or more, based on actual LWR experience. On cross-examination, however, it was established that the Nuclear Safety article analyzed the frequency of " containment failures" during leak testing for technical specification compliance in LWR's. The technical l cpecification requirements for LWR leak testing are set at substantially lower leak rate values (about a factor of 10 lower) than the design basis leak rate specified for reactor design basis accident analysis (0.1 volume % per day).      In contrast, the Staff's Appendix J analysis estimated the frequency of contain-ment breach or total loss of containment function, and not the frequency of small leaks encountered during leak testing.      Thus,
 
the data in the Nuclear S'afety article are simply inapplicable to the containment failure frequency estimated by the Staff in Appendix J. Finding 87.
l On the basis of the evidence in the record as a whole, the Board finds that the Staff has conservatively estimated the LOHS frequency, it has properly accounted for pipe rupture fre-quency in estimating the LOHS frequency, and it has conserva-tively estimated the containment failure frequency due to loss of Citigating systems.          Findings 81-87. Further, the Board finds that the Staff's Appendix J          analysis conservatively estimates the risks of beyond-design basis accidents for CRBRP, that those risks are comparable to those for current generation LWR's, and that the Staff's Appendix J analysis complies with all elements of the Commission's Statement of Interim Policy for Nuclear Power Plant Accident Considerations under the National Environmental Policy Act of 1969.          Findings 73-87.
: 2. Contention 5b) (Risks to Nearby Energy and National Security Facilities due to CRBRP Accidents)
In contention Sb), Intervenors assert that an alternate site should be chosen for CRBRP because an accident at the pro-posed CRBRP site "could" result in long-term evacuation of the Oak Ridge National Laboratory (ORNL), the Oak Ridge Gaseous Diffusion Plant (ORGDP), or the Y-12 Plant.          For the reatons i
 
which follow, the Board finds that Intervenors' contention is without merit.
Oak Ridge National Laboratory is involved in energy research and, even assuming the necessity for long-term evacua-t    tion, there would be no significant impact on either national security or national energy supply.      Findings 89-90. The Y-12 Plant is involved in DOE's weapons program and its long-term ovacuation would not significantly affect national energy supply. Finding 90. Finally, ORGDP plays no role in national security matters. Long-term evacuation of ORGDP therefore would not impact national security. Finding 90. Thus, the inquiry under contention 5b) is reduced to consideration of the risks to national security in respect to Y-12, and the risks to national energy supply in respect to ORGDP.
In considering the ef fects of a CRBRP accident on ORGDP and Y-12, both Applicants and Staff conservatively used the site suitability source term releases to calculate doses at the two facilities. The consequences of the SSST release are greater than any design basis accident for CRBRP, and thus the SSST release bounds the potential effects of credible CRBRP accidents upon the two facilities.-/    Finding 91.
I    */
Applicants and Staff made a number of additional conserva-tive assumptions which had the effect of increasing doses at the two facilities. Finding 92.
 
4
\                                                                      l l
The doses resulting from an SSST release at CRBRP are l below the DOE occupational exposure standard and the EPA Protective Action Guideline and would not lead to long-term ovacuation of either the ORGDP or the Y-12 Plant. Nor would such a release result in any curtailment of production at either facility for even a short period of time. Findings 93-94. /
As an additional measure of conservatism, both Applicants and Staff also considered the effects of an HCDA on ORGDP and the Y-12 Plant. Finding 95. In analyzing the risks of an HCDA on these facilities, both Applicants and Staff chose accident sequences involving similar containment conditions, and conservatively analyzed the consequences. Finding 97. Although HCDA's with mcce severe consequences can be postulated, the prob-l ability of such events is extremely low, and the risk (product of probabilities and consequences) of more severe HCDA's is no
  --*/  Applicants' and Staf f's analyses of the ef fects of an accident at CRBRP using SSST releases clearly meet the requirements of the National Environmental Policy Act, which requires consideration of reasonably foreseeable environ-mental impacts. Natural Resources Defense Council, Inc. v.
l      Morton, 458 F.2d 827, 836 (D.C. Cir. 1972) cited"with approval in Vermont Yankee Nuclear Power Corp. v. NRDC, 435 U.S. 519, 551 (1978); Scientists' Institute for Public Information, Inc. v. AEC (SIPI), 481 F.2d 1079, 1092 (D.C.
Cir. 1973). Consolidated Edison Co. of New York, Inc.
i (Indian Point Station, Unit No. 2), ALAB-188, 7 AEC 323, 343 l      (1974), remanded'on'other~~ grounds, CLI-74-23, 7 AEC 947; Gulf States Utilities (River Bend Station, Units 1 & 2),
LBP-75-50, 2 NRC 419, 447-48 (1975); Perkins, supra, 8 NRC at 480.
 
l greater than for the HCDA events analyzed by Applicants and Staff. Finding 96.
Assuming the extremely unlikely occurrence of an HCDA at CRBRP, there would no significant impact on the Y-12 Plant and.
long-term evacuation would not be necessary.      Finding 98.
Similarly, at ORGDP, there would not be any lonn-term evacuation and no significant impact on production would result.      Finding
: 99. /  Accordingly, even assuming an HCDA at CRBRP, there would not be any effect on national energy supply or national security.
Finding 100.
l Although Intervenors did not submit any evidence in support of this contention, Intervenors challenge the assumptions i    made by Applicants and Staff in the following respects.        First, Intervenors argue that the Staff's and Applicants' analyses are inadequate because wet deposition was not taken into account.
Intervenors ignore the fact, however, that both Applicants and Staff made the highly conservative assumption that the radioac-i tive plume would not be depleted until it reached the particular facility location. Had wet deposition been taken into account,
    */
The Staff, using highly conservative meteorological assump-tions, concluded that, in the event of an HCDA, it would be l          necessary to evacuate ORGDP. The record, however, also l          demonstrates:  (1) that the long-term evacuation of ORGDP l          would not impact national energy supply (Finding 99); (2)
I that the risk of such an accident is extremely low (Finding 103); and (3) that DOE has found these risks acceptable from a programmatic standpoint (Finding 103).
 
it would have depleted the plume, thus resulting in low.er doses. Finding 102.
Second, Intervenors contend that HCDA's with more esvere consequences should have been analyzed.                While certain HCDA events would result in more severe consequences than the HCDA events analyzed by Staff and Applicants, the probability of cuch events is extremely low, and the risk (product of
; probability and consequences) for more severe HCDA events is not greater than the risk of the HCDA events analyzed by Staff and Applicants.            Findings 96, 103.
Third, Intervenors claim that the analysis should have considered the effects of ground c,ontamination beyond seven days. In calculating ground contamination, Applicants assumed that the 30-day release was concentrated into a seven-day period.      The major constituents of deposited radionuclides are the short-lived isotopes iodine-121 and neptunium-239.                The calculated dose would decrease rapidly due to decay, and the dose contribution after seven days would not be significant.                Finding 104. Thus, Applicants' consideration of a seven-day ground contamination dose was reasonable.
l Finally, Intervenors contend that there is some unspecified uncertainty in regard to Applicants' gas sparging analysis.      There is no evidence of any uncertainty in regard to I the gas sparging analysis.                Rather, the record indicates that 4
      ~    --    - - - -      _ , - - ,-,
 
Applicants' gas sparging analysis is conservative.                      Footnote accompanying Finding 104.
In summary, the record clearly demonstrates that there would be no significant risk to either national energy supply or national security in the highly unlikely event of a severe accident at CRBRP.
: 3. Cont en ti ons" 4"and" 6b ) 4 )" ( Sa'f eqcu rds' ' I mpa ct s )
In Contentions 4 and 6b)4), Intervenors broadly challenge both the Applicants' and Staff's analyses of safeguards at CRBRP and its supporting fuel cycle facilities.                      In consider-ing Intervenors' safeguards contentions, the Board is not called upon to address the adequacy of safeguards at CRBRP.                          The ques-tion of adequacy of safeguards is appropriately deferred until the operating license stage of this proceeding. /                      In addition, the supporting fuel cycle facilities are subject to DOE's exclu-cive jurisdiction, thus circumscribing our review as to adequacy of safeguards for those facilities.
At this stage of the proceeding the scope of our review                                I is limited. As we stated at the time we admitted Intervenors' cafeguards contentions:                                                                          I I
We hold that an evaluation of the potential                                            l cost of safeguarding the CRBR, fuel cycle                                              j facilities and transportation supports should                                          l
*/
-~
E.9., Power Authority of the State of New York (Greene County Nuclear Power Plant), LBP-79-8, 9 NRC 339 (1979).
 
be included in the NEPA cost benefit analysis.  ...
Board Order dated April 6, 1976 at 9.                      Thus, contentions 4 and 6b)4) are limited to consideration of the influence of safeguards costs on the NEPA cost / benefit balance.
As Intervenors concede, there are few, if any, factual disputes between the parties.      Due to the inherent design characteristics of CRBRP, and in particular, its fuel handling system, theft of plutonium is a highly unlikely event irrespec-tive of the physical security system.                      Findings 110-111.
Similarly, the design of the plant protection system, control system and hardware make radiological' sabotage a highly unlikely cvent. Finding 112.                                            -
There is no dispute that the safeguards system planned to be implemented at CRBRP will likely meet or exceed all NRC requirements and provide a high degree of protection against acts of sabotage and theft. The planned CRBRP safeguards system will cmploy the latest advances in safeguards technology, including the use of explosives detectors, metal detectors, dual computer-based card readers, photo identification systems and redundant communications systems. Findings 113-121.
The capital cost of the safeguards system for CRBRP will be approximately $3.8 million, or less than one percent of total plant costs. In light of the modular design of the system, the cost is not likely to increase significantly in the future.
 
l                                                                                      l Finding 124. The total annual cost of operating the safeguards system is estimated at $2.5 million.            Finding 124.
In light of the high degree of inherent protection af forded by the CRBRP design, and upon consideration of the protection afforded by the safeguards system, we find that theft or radiological sabotage are extremely unlikely events, and the risks of such acts are not significant.              Moreover, the economic costs of safeguarding CRBRP against theft or sabotage are, and are likely to remain, small fractions of the tetal plant cost.
Finding 125. We conclude, therefore, that the safeguards system for CRBRP will have little or no impact on the cost / benefit balance for CRBRP.
The CRBRP fuel cycle includes mixed oxide fuel fabrica-tion, blanket element fabrication, reprocessing, management of wastes generated by the various facilities, and transportation of fuel and wastes among the various facilities.                    Findings 126-136.
The DOE fabrication and reprocessing facilities and transportation system are not licensed by NRC or subject to NRC safeguards requirements.            The DOE threat guidance, howo~ar, is comparable to the design basis threats embodied in the eciacing NRC safeguards regulations.            Finding 107.          Safeguards designed in accordance with DOE requirements will provide a level of protec-tion against theft or sabotage at least as high as that provided by safeguards designed in accordance with NRC requirements.
Findings 106-107.
 
As with the CRBRP safeguards, there are no factual disputes regarding the safeguards which will be implemented within the CRBRP fuel cycle.                                            The DOE fuel fabrication facility and the Developmental Reprocessing Plant (DEP) will. implement i
physical security and material control and accounting systems which will make theft or radiological sabotage highly unlikely events. Findings 127-135.                                            This can be accomplished using existing safeguards technologies which are routinely used in existing facilities, or advanced technologies which have been successfully tested on a demonstration scale.                                              Findings 127-128, 132-134.
I The DOE Transportation Safeguards System, which is planned for use in transporting fresh as well as spent CRBRP fuel, is presently used extensively to transport special nuclear material and other sensitive material throughout the United States and provides weapons level protection to all such shipments.                The system includes specially designed transportation squipment, trained couriers and a nationwide communications system, which provide a high degree of protection against theft or sabotage.                            Finding 136.
The wastes generated by CRBRP and its fuel cycle facilities are not attractive theft or sabotage targets due to their high radioactivity and low concentrations of plutonium.
Findings 141-143.                                            The wastes will be stored at a federal 0
              - - - - - . _ - _ _ _ - . _ _ _ . -- - - - - - -              .,-.n.- --. -
 
l i
repository which will be licensed by NRC and thus subject to NRC cafeguards requirements.          Finding 143.
The total safeguards capital costs associated with the CRBRP fuel cycle amount to only $5.5 million, while total annual operating costs are $3.1 million.            Findings 129, 135, 137, 140, 144. These costs represent only a small fraction of total plant costs and would not significantly affect the cost / benefit balance. Theft or sabotage are highly unlikely events, and the risks associated with safeguarding fuel cycle facilities are extremely low.          Findings 126-143. Thus the environmental impacts of safeguarding the CRBRP fuel cycle are expectsd to be insignificant.
Although Intervenors did not submit the testimony of a qualified safeguards expert (Finding 146), Intervenors did raise a number of issues in challenging Applicants' and Staff's analyses of the environmental effects of safeguarding CRBRP and its fuel cycle facilities.
First, based on empirical evidence, Intervenors claim that theft or sabotage is " credible".          In making this assertion, however, Intervenors ignored the safeguards measures which will ba implemented at CRBRP and its fuel cycle facilities, and thus failed to provide any nexus between their empirical evidence and the CRBRP fuel cycle facilities.            Intervenors' evidence consisted of:  1) actions directed against plants under construction which did not have any safeguards systems (Findings 151, 152, 154); 2)
 
actions directed at facilities which did not handle Special 1
Nuclear Material (Findings 149, 150); or 3) actions directed at l
facilities which did not have safeguards requirements as stringent as those which will imposed by NRC and DOE for CRBRP fuel cycle facilities (Findings 148, 149, 150). In short, Intervenors' empirical evidence fails to establish that success-ful sabotage or thef t at CRBRP or its fuel cycle facilities are credible events.
Intervenors argue that the analysis of the fuel cycle cafeguards is inadequate because alternatives to the DRP were not considered. In fact, a number of reprocessing alternatives were considered, including the DRP. Findings 130-131, 155. The DRP, however, represented the highest safeguards cost among the three  ,
1 reprocessing alternatives considered in the Applicants' and Staff's analyses, and thus the DRP costs bound the costs of other  )
alternatives. Findings 131, 155.                                !
Intervenors contend that the Staff's environmental analysis is inadequate because possible civil liberties viola-tions which might occur in the event of successful Jseft or cabotage were not considered. The record evidence, however,
;  clearly establishes that such risks are no greater than those already encountered and accepted in the various existing military programs. In addition, in light of the fact that theft or sabo-tage at CRBRP and its fuel cycle facilities is highly unlikely (see Findings 106-145), the possibility of civil liberties
 
violations is even less likely and represents an insignificant incremental risk. Finally, there is nothing in the record to
,  suggest that such civil liberties violations have ever occurred cither in the military programs or in the case of commercial reactors. Finding 156. In short, the likelihood of such violations is slight.
Intervenors have questioned whether Applicants, including DOE, will in fact implement the safeguards systems i
dnscribed during the LWA hearings. The Board, having heard and questioned the Applicants' witnesses, believes that the Appli-cants, including DOE, are strongly committed to implementation of ef fective safeguards systems for CRBRP and its fuel cycle facilities. Aside from Applicants' commitments, NRC regulations and DOE orders mandate implementation of effective safeguards, thus providing additional assurance that effective safeguards will be implemented. Finding 157.
Finally, Intervenors argue that small quantities of fresh mixed oxide fuel can be converted into a clandestine fission explosive (CFE). Contrary to Intervenors' assertion,it j  is simply not true that small quantities of fresh fuel could be otolen and used to construct a CFE. In order to obtain the 6-12 kilograms of plutonium to construct a CFE, it would be necessary l
to steal three fuel assemblies, each of which is 14 feet long and weighs 450 pounds. Finding 159. No evidence of any kind was presented by Intervenors that a theft of this magnitude is even
 
                                                        - 62'-
remotely possible. Even assuming, however, the highly unlikely theft of three fuel assemblies weighing a total of 1350 pounds, construction of a CFE would be highly unlikely even for a person with a nuclear physics or nuclear engineering background, and then it would depend to a large degree on luck.                            Finding 159.
In conclusion, the Board finds that Intervenors' contentions regarding safeguards at CRBRP are without merit. The enfeguards planned to be implemented at CRBRP and its fuel cycle facilities will not result in any significant environmental impacts and the costs do not significantly affect the cost /
benefit balance.
: 4. Contentions ~6b)l) and'6b)3)'(Fuel' Cycle Impacts)
Intervenors' contentions 6b)l) and 6b)3) allege that Applicants and Staff have not adequately analyzed the environ-m:ntal impacts of the CRBRP fuel cycle with regard to fuel fabri-cation, fuel reprocessing, and disposal of wastes.                            During the course of the proceedings, Intervenors advanced four arguments in support of these contentions:                          a)  the isotopic concentrations of Pu-238 and Pu-241 in the spent fuel from the CRBRP may be under-cstimated, thereby underestimating the plutonium doses from reprocessing operations by a factor of 2 to 4.3; b) the environ-nental risks from fuel reprocessing were not conservatively satimated because alternative reprocessing facilities were not o
considered; c) containment factors for fuel cycle facilities were
 
not adequater and d) environmental impacts from CRBRP waste nanagement were not properly evaluated.
Applicants' and Staff's analyses of reprocessing impacts were based on isotopic concentrations of Pu-238 and Pu-241 which exceeded those expected for the CRBRP fuel cycle, and thus the calculated doses exceeded the expected contribution of plutonium doses. Finding 170. There are ample quantities of low burn-up LWR fuel with isotopic concentrations below those assumed in the Applicants' and Staff's analyses.      Furthermore, if fuel with higher isotopic concentrations than aasumed in the analysis were proposed for use in the future, such use would be cubject to Commission review and approval.      Finding 171.
Even if higher burn-up fuel were used, it would cause no significant changes in environmental effects.      All plutonium isotopes, including Pu-238 and Pu-241, account for approximately one-tenth of one percent of the total dose from CRBRP fuel cycle operations. Even if those doses were increased by a factor of 2 to 4.3, the overall effect would be insignificant.      Finding 172.
Therefore, based on the above considerations, the Board finds that there is no basis for projecting any change in the isotopic composition of the spent fuel analyzed by Applicants and Staff.
Furthermore, even if the spent fuel isotopic concentration were changed and'the Pu-238 and Pu-241 source terms did increase by a factor of 2 to 4.3, the Board finds that the resulting increase in environmental impact would not be significant.
 
l                                                                                            1 l                                        .
Applicants' analysis of fuel reprocessing impacts was based on utilization of the projected Developmental Reprocessing Plant. The Staff independently evaluated the likely environ-nental impacts of fuel reprocessing based on utilization of the DRP. The Staff's analysis for DRP assumed the total release of i
all tritium (H-3) and all carbon-14 (C-14) during reprocessing.
j      Those two elements dominate radiological impacts, accounting for over 99% of the estimated dose. These releases could not be exceeded no matter what alternative reprocessing facility was considered. /    Findings 173-177. Therefore, the Board finds that the Staff and Applicants' analyses of the environmental impacts of fuel reprocessing conservatively bound the environmental impacts which could be expected for any alternative reprocessing facility.
Applicants' and Staff's expert witnesses presented probative, reliable evidence to support the validity of the containment factors for the CRBRP fuel fabrication and repro-cessing facilities used in their analyses of radiological releases. Intervenors' witness Cochran presented calculations to show that, based on operational experience, the actual contain-ment factor values should be a factor of ten worse.                    On cross-oxamination it was established that these calculations were l
        --*/  The Staff's conservative assumptions of total releases of H-3 and C-14 overestimate the expected doses by a factor of about five. Finding 176.
l
 
unreliable and admittedly incomplete. Finding 180. The con-tainment factors used by Staff and Applicants were based on a substantial body of experience in the design and use of high-officiency particulate absolute (HEPA) filters, and in any event, the simple addition of only one bank of HEPA filters would improve performance by a factor of one thousand. Findings 178-182.
4 Intervenors' witness Johnson also argued that the radiological releases fom CRBRP fuel cycle facilities have been underestimated based on experience at Rocky Flats. The evidence J
in the record, however, demonstrates that the Rocky Flats facility is sufficiently dissimilar from the CRBRP fuel cycle facilities in terms of function, process, products and releases, so that any attempt to use Rocky Flats as the basis for analysis of expected CRBRP fuel cycle impacts is meaningless. Findings 182-184. Based on the evidence presented, the Board finds that the containment factors assessed in the Staff's and Applicants' analyses are reasonable and can be achieved.
The Staff's analysis of potential health effects asso-ciated with high level waste disposal was based on proposed EPA otandards which contain limits based upon an upper bound value of 1,000 health effects over the first 10,000 years after closure.
CRBRP wastes would occupy 0.36-1.0 percent of the total reposi-tory capacity. Thus, a maximum of 3.6-10 health effects over 10,000 years can be attributed to management of CRBRP high level waste. Findings 185-189.
1 i
 
Intervenors incorrectly applied the proposed EPA standard and projected all health effects from CRBRP waste canagement as occurring over 30 years, not 10,000 years, thus yielding 0.3 health effects per year.      In actuality, the correct projection would be 0.00036-0.001 health effects per year, and even the upper bound value would yield an insignificant effect.
Finding 188.
Based on the evidence presented, the Board finds that the environmental impacts associated with the CRBRP fuel cycle have been conservatively estimated, and are not significant.
: 5. Contentions'5a)''and''7c)''(Alternative Sites)
Intervenors' contentions Sa) and 7c) questioned the adequacy of the Applicants' and Staff's analysis of alternatives l
to the Clinch River site. The Commission's decision in United States Energy Research'and' Development Administrati'on (Clinch River Breeder Reactor Plant), CLI-76-13, 4 NRC 67 (1976) estab-lished certain principles for the conduct of this proceeding and the review and consideration of alternatives.      These include:
a) "the need for a liquid metal fast breeder reactor program, including its objective, structure, and timing" shall be taken as established; b) "the need for a demonstration-scale facility to test the feasibility of liquid metal fast breeder reactors when l  operated as part of the power generation facilities of an elec-l tric utility system, including its timing and objectives" shall l
 
be taken as established; c) alternatives for meeting the objec-tives "are to be evaluated in terms of the objectives defined in the (programmatic] impact statement," and d) " consideration of' alternatives need go no further then to establish whether or not substantially better alternatives are likely to be available."
Applicants and Staff presented extensive testimony and documentary evidence in support of their judgment that no sub-stantially better siting alternative exists, while Intervenors presented no evidence on these contentions.
There is no dispute that the alternative siting concepts of a hook-on plant, co-location, and underground siting were considered and properly rejected as not substantially better than an all-new plant at the Clinch River site. Finding 195. A broad range of alternative sites were considered, including sites:  1) within the TVA power service area, 2) outside the TVA service area but owned by TVA, and 3) owned by DOE throughout the 1
l United States. The record shows that none of the alternative l
sites considered are substantially better than the CRBRP site.
Nono were preferable on the basis of environmental factors.
Findings 196-201. In addition, none were better than the Clinch River site from the standpoint of cost and programmatic factors. /  Findings 206-209.
  --*/  In the oral argument, Intervenors' counsel cited two cases to support their proposition that Applicants' and Staff's evaluations of alternative sites were inedequate. Both cases, however are completely unrelated to the facts (Continued)
 
I Intervenors argued that the Clinch River site must be rejected because other sites had more favorable atmospheric diffusion and population density characteristics.                    As the l  uncontradicted evidence clearly shows, the atmospheric diffusion characteristics of the Clinch River site are comparable to the alternative TVA sites.                While the meteorology for the alternative DOE sites is somewhat more favorable, the Clinch River site is acceptable for a nuclear facility.                  Finding 203. Any differences in population density between Clinch River and the alternative
                                                      */
oites are similarly insignificant.-                    Finding 204.
presently before this Board. In Florida Power and Light Company (St. Lucie Nuclear Power Plant, Unit No. 2), ALAB-335, 3 NRC 830 (1976), the Staff's environmental analysis examined no specific alternative sites but rather compared the proposed site with two hypothetical sites representing composites of characteristics which would typify a coastal site and an inland site. In Boston Edison Company (Pilgrim Nuclear Generating Station, Unit 2), ALAB-479, 7 NRC 774 (1978) the Staff analysis again did not consider specific alternative sitis but rather " lumped" sites into three gen-              i eral types or categories -- inland, coastal and offshore --
and compared the proposed site with those categories. I'd.
at 784. The Appeal Board in both cases found the analysis inadequate because of the Staff's failure to consider specific alternative sites. In contrast, in the present proceeding both Applicants and Staff have looked in depth at a broad range of specific sites.
    -*/      The FY 1980 NRC Authorization Act does not change this result. Pub. L. No. 96-286, 94 Stat. 780 (June 30, 1980).
Section 108, which contemplated consideration of population density independent of design, expressly exempted any facil-ity "for which an application for a construction permit was filed on or before October 1, 1979." The CRBRP application was filed on October 11, 1974. Similarly, NRC's Notice of Proposed Rulemaking related to Section 108 exempted such facilities. 45 Fed. Reg. 50350 (July 29, 1980).              In connec-tion with Congress' consideration of the FY 1983 Act, it was
(              established that the FY 1980 Act had expired, (128 Cong.
l (Continued) l l
 
Differences in site meteorological and population density characteristics are important only to the extent that such differences might present significant differences in radiological risk. CRBRP will meet the site suitability dose guidelines for any design basis accident. Findings 23-24. The effects of routine releases at the Clinch River site are already eo small that a further reduction in risk would not be signi-ficant. Findings 244-258. The risk of beyond-design basis accidents is similarly low and comparable to that of LWR's.      Any further reduction in the residual risk from severe accidents would not constitute a si.gnificant reduction in risk. Findings 22, 73-87. Accordingly, from the standpoint of radiological risk, the alternative sites are not substantially better than the Clinch River site. Finding 205.
Based on consideration of all the evidence presented, the Board finds that there are no alternative sites which are substantially better alternatives to the Clinch River site for meeting programmatic objectives.
1 Rec. S.13056 (October 1, 1982)), and although the Congress had expressed support for the policies and objectives of the FY 1980 Act (128 Cong. Rec. H. 7677 (September 28, 1982)),
there is nothing in the legislative history of the 1983 Act which would impose new population density requirements for any facility, much less those expressly exempted from the 1980 Act.
: 6. Contentions 7a) and 7b) (Likelihood of Meeting Programmatic Objectives'and' Design' Alternatives)(
Intervenors' raised three basic issues regarding programmatic objectives and design alternatives:    a)  whether the CRBRP is reasonably likely to meet its programmatic objectives (contentions 7a)1) and 3)); b)    whether the CRBRP will provide information relevant to commercial size LMFBR's (contention 7a)2)); and c)    whether the informational objectives of the CRBRP might be substantially better satisfied by alternative design features different from those in the CRBRP (contention 7b)).
Findings 210-211.
Intervenors presented no evidence regarding these contentions, instead limiting themselves to cross-examination of Staff's and Applicants' witnesses.      The uncontradicted evidence in the record shows that the CRBRP is reasonably likely to achieve all of its programmatic objectives (Findings 213-225) in a timely manner /    (Finding 227). In this regard, Intervenors did question as to whether constructing and testing a precise t
  */
  -~
The Intervenors questioned Applicants' witnesses at length about the timing objective with the purpose, as Intervenors' I
counsel finally admitted, of demonstrating that the timing objective is meaningless. Finding 227 and accompanying footnote. Since the Commission has ruled that the validity
,        of the timing objective is not open to question in this proceeding, Cli'nch' Rive'r, 4 NRC at 92, this Board need not consider the validity of Intervenors' contention as to the meaningfulness of the timing objective. Suffice it to say that the CRBRP timing objective, when considered with the other project objectives, is likely to be achieved.
Findings 213-215.
l
 
I i
i prototype of the CRBRP steam generator would be a better t
l      alternative for achieving the technical performance objective.
Applicants and Staff clearly demonstrated, however, that the steam generator design and verification test program are well founded and that the technical risk of an undetected major design defect, requiring redesign and delay after installation, is very remote.        The assurances gained from testing a precise prototype cannot technically justify the delay required to construct such a prototype.          Findings 217 - 218.
Reliable, probative, uncontradicted evidence was also presented to demonstrate that the CRBRP would provide information which is relevant and of substantial value to the construction and operation of commarcial LMFBR's.                          Findings 228-234.                  Appli-
!    cants' and Staff's witnesses also demonstrated that there are no known alternative design features which might substantially better satisfy the informational objectives of the LMFBR program.
Intervenors raised no substantial questions to the contrary. /
      */
    -~
Intervenors questioned Applicants' witnesses as to whether a no-vent containment would be a substantially betcar alter-native. Applicants' witness Kaushal, relying upon t
Applicants' exhibits, demonstrated that venting during normal operations improves plant operability and maintain-ability. For a design basis accident, the containment is isolated and unvented. In the event of a beyond-design basis Hypothetical Core Disruptive Accident the filtered controlled vent capability assures that containment integrity cannot be challenged, while maintaining radio-logical releases at an acceptably low level. One cannot, as Inter enors' argument suggests, design a perfect contain-ment, and the vent capability provides a prudent additional margin of assurance that the risks of severe accidents will (Continued) l
 
Based on the evidence presented, the Board concludes that:      1) the CRBRP is reasonably likely to meet its programmatic objectives; 2) the CRBRP will provide information relevant and of substantial value to commercial LMFBR's; and 3) there are no cubstantially better design alternatives for satisfying the informational objectives of the CRBRP.                  Findings 210-243.
: 7. Contentions lib) and lic) (Genetic and Somatic Tffects"of'CRBBP Operation)'
In Contentions lib) and lic), Intervenors assert that neither Staff nor Applicants adequately analyzed the possible genetic or somatic health effects which might occur as a result of normal operation of CRBRP.                  Intervenors did not, however introduce any evidence in support of these contentions.                    In contrast, both Applicants and Staff introduced testimony of highly qualified experts in the fields of genetics and radiation health effects.                The uncontradicted evidence demonstrates that the genetic and somatic effects of CRBRP operation will not be significant.
In estimating genetic effects, both Applicants' and Staff's experts used accepted methodologies and employed a number of conservative assumptions.                  Findings 246, 252. Based on these methodologies and assumptions, these experts calculated genetic be acceptably low. Therefore, a no-vent containment is not a substantially better alternative. Finding 242.
 
l l
l cffects which would be virtually undetectable in light of the current incidence of genetic effects occurring in the popula-tion. Staff's witness Dr. Bender, for example, who served on the BEIR-III Committee, calculated the combined increase in risk to the general public and the occupational work force as amounting at most to 0.00002 percent for the first generation and a lesser risk in subsequent generations. Finding 252.
Similarly, in estimating the somatic effects both Applicants and Staff employed accepted methodologies and used conservative assumptions. The estimated increase in somatic effects to the general public range from 0.000015 to 0.00005 per reactor year. The estimated effects among the workers range from 0.07 to 0.2 per reactor year. Finding 251. Moreover, the lower range of these estimates could in fact be zero. Finding 251. In contrast, the current incidence of cancer in the general population is sixteen percent. Finding 251.
Although not disputing either Applicants' or Staff's estimates, Intervenors claim that:    1) uncertainties in the BEIR-III Report were not taken into account;    2) the views of other experts were not taken into account; and 3) recent reevaluation of the neutron and gamma doses at Nagasaki and Hiroshima were not considered.
As to the uncertainties in the BEIR-III Report, the uncontradicted record evidenca demonstrates that the values calculated by Applicants and Staff in considering genetic and
 
comatic health effects were upper bound limits.          Both Applicants and Staff made conservative assumptions which had the effect of overestimating the expected health effects from operation of CRB RP . Finding 256.
As to the views of experts, both Applicants and Staff relied upon well qualified experts in the fields of genetic and comatic health effects due to exposure to radiation.          In per-forming their analyses, these experts used data and metholodogies which are widely accepted in the scientific community.          In addition, these experts considered the contrary views of Gofman,
                        */
among other experts.-    Dr. Bender, an expert in genetics, noted that he did not consider Gofman an expert in genetics and that Gofman's analysis reflects a misunderstanding of certain genetic principles. Applicants' testimony addressed Gofman's views on genetic effects, and showed that several of Gofman's hypotheses concerning genetic effects had no merit.          Finding 257.
Finally, the recent reevaluation of the neutron and gamma doses at Nagasaki and Hiroshima does not create any substantial uncertainty with regard to the analysis of health offects. The record stands uncontradicted that substantial l
l l
  */    See also Findings 58-60, supra.
 
changes in the BEIR-III Report risk estimators are not expected as a result of the reevaluation. Finding 258.-/
On the basis of the evidence in the record as whole, the Board finds that Applicants' and Staff's analyses of the genetic and somatic ef fects of CRBRP operation overestimate the oxpected effects, and even on that basis, those effects are not significant.
C. ULTIMATE' LEGAL ISSUE Contrary to Intervenors' argument, neither the language of the LWA regulation, 10 C.F.R. $ 50.10(e), nor its rulemaking history, nor any Commission precedent contain authority for the proposition that it cannot be applied to the CRBRP. Rather, l
: 1) the provisions of the Atomic Energy Act; and 2) the rulemaking history of Section 50.10(e) conclusively establish the applic-ability of the LWA procedures to CRBRP. Further, on the basis of the evidence in the record as a whole, we find that the LWA procedure can be applied to CRBRP. Opinion, Section A.1, supra; Findings of Fact and Conclusions of Law, infra.
i 1
  */
  ~~
For the most part, Intervenors' questions regarding the reevaluation of the Nagasaki and Hiroshima data were directed to the linear quadratic model in BEIR-III. This model was not used by either Applicants or Staff. See Finding 258, first footnote.
: 1. The Atomic Energy Act Requires That Section'50'lO(e)~Be'Available'To'The'CRBRP Under the Atomic Energy Act, 42 U.S.C. 2011 et seq.,
the Energy Reorganization Act, 42 U.S.C. 5801 et seq.', and the CRBRP Authorization Acts, Pub. L. No. 91-273, as amended, the NRC regulatory requirements, including 10 C.F.R. $ 50.10(e), are fully applicable to CRBRP.          Section 202(1) of the Energy Reorganization Act provides that NRC shall have licensing and regulatory authority as to:
Demonstration Liquid Metal Fast Breeder Reactors when operated as part of the power generation facilities of an electric utility system, or when operated in any other manner for the purpose of demonstrating the suitability for commercial application of such a reactor.
As presently defined in the authorizing legislation, Pub. L. No.
91-273, ajs acended, the CRBRP project is subject to NRC licensing l  and regulatory jurisdiction.          The statutory criteria and program justification data, which were approved by the authorizing legislation, contemplate operation of the facility as part of the
                                                            */
power generation facilities of a utility system.-            Consequently, the project falls within the ambit of Section 202(1) of the Energy Reorganization Act.          Further, since the project is
  */
  -~
See Pub. L. No. 94-187; ERDA Authorizing Legislation for TIscal Year 1976:        Hearings on Fission Power Reactor Development Before Subcomm. on Legislation of the Joint Comm. on Atomic Energy, 94th Cong., 1st Sess., Part 4 at 2280-87 (1975) (Letter of Robert C. Seamans, Jr.)
[ hereinafter, 1975 JCAE Hearings].
 
authorized as the fourth round of the Co-operative Power Reactor Demonstration Program, / any license applied for would be issued
                                                        ** /
under Section 104(b) of the Atomic Energy Act.--
The Commission must apply the same procedurer for Sec-i tion 104(b) licenses as are available for Section 103 commercial nuclear power plant licenses.      Section 104(b) provides that:
[i]n issuing licenses under this subsection, the Commission shall impose the minimum am'ount"of"such rwgu1~at' ions"and' terms of license as will permit the commission to fulfill its obligations under this chapter.
42 U.S.C. 2134(b) (emphasis added).        S. Rep. No. 99, 83d Cong.,
2d. Sess. (1954), which accompanied this legislation, explained that the Commission should apply this provision so as to impose only those regulations on demonstration plants that would be compatible with those later imposed on a commercial plant of the came type if the technology were late- to prove commercially feasible. S. Rep. No. 1699, reprinted"in U.S. Code Cong. & Ad.
l News 3475 (1954).      In other words, the regulatory requirements i
for Section 103 commercial nuclear power plant licenses form a ceiling for the requirements imposed upon Section 104(b) licenses. Thus, the Commission cannot impose more stringent l procedures on Section 104(b) reactors than it imposes on Section 103 reactors. Similarly, whatever procedures are generally See Section 106 of Pub. L. No. 91-273, as amended; see 1975
  -*/    TCXE Hearings at 2280.
  **/    See 42 U.S.C. 2134b.; 42 U.S.C.      2132c.
 
available to commercial applicants must be available to Section 104(b) applicants.      Inasmuch as Section 50.10(e) =akes no distinction between classes of reactors, it must be applicable to Section 104(b), as well as to Section 103 commercial reactors.
: 2. The' ~ Rulemaking ' Hi s t~o ry"o f' ' S ect ion" 50 710T e )
The history of 10 C.F.R. { 50.10(e) provides persuasive authority for the proposition that it applies to CRBRP.                                        Prior to the enactment of the National Environmental Policy Act of 1969 (NEPA), 42 U.S.C. } 4321 e't seq., applicants for a construction permit routinely conducted non-safety related site preparation activities without any prior Commission review or authorization. Thus, prior to passage of NEPA, CRBRP could have begun site preparation without Commission approval, whether or not it is a first-of-a-kind reactor.
With the enactment of NEPA and the decision in Calvert Clif f s
* Coordinating 'Committ'ee"v.' AEC, 449 F.2d 1109 (D.C. Cir.
1971) the applicable regulations were changed to prohibit site preparation work which could adversely affect the environment prior to the issuance of a construction permit.                                    Subsequently, the present LWA regulation was proposed and adopted.                                        In promulgating Section 50.10(e), the commission made the following statement:
Prior to the enactment of the National Environmental Policy Act of 1969 (NEPA) and the amendments to } 50.10 adopted by the Commission on March 21, 1972 (37 F.R. 5745),
site excavation for safety related structures l
I
 
was generally undertaken by Applicants without any prior Commission review. The essential distinction between the past situation and the present one is that NEPA now applies to certain Commission actions.
However this essential difference is accommodated in the amendments by the requirement that there be a full NEPA review and hearing on NEPA issues covered by the Commission's NEPA regulations prior to authorizing any on-site work otherwise generally prohibited by 50.10(c). The approach in the instant rule ... i's consist ~ent ' with 'the"past ' practice.
39 Fed. Reg. 14506, 14507 (April 24, 1974)
In its present form, 10 C.F.R.        $ 50.10(e) allows essen-tially the same scope of site preparation work that was permis-oible prior to the enactment of NEPA, but after the completion of the NEPA review required for a construction permit.          Thus, while the promulgation of 10 C.F.R. $ 50.10(e) postponed commencement of site preparation work until after completion of NEPA reviews for nuclear reactors, it retained past Commission practice as to the scope of allowable pre-construction activities, and effected no change in regard to whether first-of-a-kind reactors could commence such activities.        Thus, the Applicants for CRBRP are cntitled to use the LWA procedures.          Indeed, any other conclusion would be inconsistent with past commission practice.
The history of 10 C.F.R.          50.10(e) also indicates that in promulgating the regulation, NRC purposely avoided prohibiting its application to first-of-a-kind plants.          Upon issuance of the proposed version of 10 C.F.R. { 50.10(e), the Commission received comments specifically urging it not to apply the proposed LWA
 
rules to new or novel designs, including the fast breeder reactor .- /  In spite of this comment, the final version of the      -
regulation did not limit the applicability of the LWA procedures in the case of new or novel designs in general, or the fast breeder reactor in particular.      Thus, the Commission has previously rejected the thrust of Intervenors' assertion in the rulemaking proceedings leading to promulgation of 10 C.F.R.
$ 50.10(e).
In summary, Intervenors' claim that 10 C.F.R. Section 50.10(e) is somehow inapplicable to CRBRP is incorrect as a matter of law. The provisions of the Energy Reorganization Act, the Atomic Energy Act and the CRBRP Authorization Acts, and the rulemaking history of 10 C.F.R.      50.10(e), conclusively estab-lish the applicability of 10 C.F.R. { 50 lO(e).        Further, on the basis of the evidence in the record as a whole, we find that the LWA procedure can be applied to CRBRP.      Opinion, Section A.l.,
supra; Findings of Facts and Conclusions of Law, infra.
IV. CONCLUSTON On the basis of the evidence of record, the Board concludes that there is reasonable assurance that the CRBRP site is suitable for a reactor of the general size and type proposed
-*/  Letter from State of New Jersey, Department of Environmental Protection to AEC, March 26, 1974.      (On file in the NRC Public Document Room, Washington, D.C., in file entitled "10      i CFR Parts 2, 50 (39 F.R. 4582) (pre-construction                  '
activity)").                                                      l l
 
i l
l frem the standpoint of radiological health and safety considera-tions, and upon consideration of all matters arising within the scope of NEPA, 42 U.S.C. $ 4321 et seq.', and 10 C.F.R. Part 51, the action called for is construction and operation of the CRBRP.
The matters examined during the evidentiary hearing which are not discussed in this opinion were considered by the Board and found either to be without merit or not to affect our decision herein.      Findings of fact and conclusions of law which are annexed hereto are incorporated in the opinion by reference as if set forth at length.      In preparing its findings of fact and conclusions of law, the Board reviewed and considered the entire record, the closing arguments of the parties, and the findings of fact proposed by the parties.      Those proposed findings not incorporated directly or inferentially in this Partial Initial Decision are rejected as being unsupported by the record of the case or as being unnecessary to the rendering of this decision.
The Board, having considered and decided all matters in controversy among the parties and having made the findings and determinations required by 10 C.F.R. $ 50.10(e), concludes that the Director of Nuclear Reactor Regulation, upon making requisite findings with respect to matters not embraced in this Partial i Initial Decision, may authorize the Applicants to conduct site preparation activities pursuant to 10 C.F.R.      50.10(e)(1) for the Clinch River Breeder Reactor Plant.
          .__ .                  ~
 
                                                                  \
;                                                                V. ORDER Wherefore, it is ordered that the Director of Nuclear Reactor Regulation is authorized, upon making requisite findings with respect to matters not embraced in this Partial Initial Decision, in accordance with the Commission's regulations, to issue to Applicants authorization to conduct site preparation activities for the Clinch River Breeder Reactor Plant pursuant to I
10 C.F.R. { 50.10(e)(1). Such authorization may be in such form and content as is appropriate in light of such findings, provided that such authorization is consistent with the conclusions of the ~
Board herein.
It is further ordered that this Partial Initial Deci-sion shall constitute the final action of the Commission forty-five (45) days after the issuance thereof, subject to any review pursuant to 10 C.F.R. (( 2.760, 2.762, 2.764, 2.785, and 2.786.
Exceptions to this Partial Initial Decision may be filed within ten (10) days after its service.        A brief in support of the exceptions shall be filed within thirty (30) days there-after and forty (40) days in the case of the Staff.        Within thirty (30) days of the filing and service of the brief of the Appellant, and forty (40) days in the case of the Staff, any I"
 
other party may file a brief'in support of, or in opposition to,  l the exceptions.
IT IS SO ORDERED.
FOR THE ATOMIC SAFETY AND LICENSING BOARD Marsnall E. Miller, Gnairman ADMINISTRATIVE JUDGE Dated at Bethesda, Maryland              '    '
This                      1983 day of        ,          Gustave A. Linenburger, Jr.
ADMINISTRATIVE JUDGE Dr. cadet H. Hand, Jr.
ADMINISTRATIVE JUDGE I
 
s WHEREFORE, on the basis of the foregoing and the annexed Proposed Findings of Fact and Conclusions of Law, Applicants respectfully request that the Board adopt this Proposed Decision.
Respectfully submitted, Of Guerge L.      rV' -
Attorney Project Management Corporation Abf  ,
William D. Luck Attorney for United States Department of Energy DATED:  January 24, 1983 OF COUNSEL:
* Thomas A. Schmutz                                                                Herbert S. Sanger, Jr.
Frank K. Peterson                                                                Lewis E. Wallace Morgan, Lewis & Bockius                                                          W. Walter LaRoche 1800 M Street, N.W.                                                              James F. Burger Washington, D. C. ?CO36                                                          Edward J. Vigluicci I i Tennessee Valley Authority Leon Silterstrom                                                                  400 West Summit Hill Drive Warren E. Bergholz, Jr.                                                          Knoxville, Tennessee 37902 U.S. Department of Energy Office of General Counsel 1000 Independence Avenue, S.W.
Room 6B-256 Washington, D. C.                                                20585
 
                                                  - F                                            VI. FINDINGS ' OF ' FACT
: 1. The Applicants in this Administrative proceeding are the United States Department of Energy (DOE), Project Management Corporation (PMC), and the Tennessee Valley Authority (TVA).            S Exh 8 at 1-1.
: 2. The other parties to the proceeding are the Staff of the United States Nuclear Regulatory Commission (Staff or NRC Staff), and the Natural Resources Defense Council (NRDC) and the Sierra Club (Intervenors).                    The Tennessee Attorney General and City of Oak Ridge have participated as an interested state and municipality, respectively, under 10 C.F.R.        I 2.715(c).
: 3. Applicants seek a Construction Permit (CP) and Operating License (OL) for a term of 40 years, pursuant to Section 104(b) of the Atomic Energy Act, for the Clinch River Breeder Reactor Plant ( CRBRP ) .                S Exh 7 at 1-2. These findings coacern the portions of the CP application which relate 90 grant or denial of a Limited Work Authorization (LWA) pursuant to 10 C.F.R.                  $ 50.lO(e)(2).
: 4. The CRBRP is a Liquid Metal Fast Breeder Reactor (LMFBR) demonstration plant with a rated output of 350 megawatts l
electrical which is proposed for location on the Clinch River in Oak Ridge, Tennessee.                    S Exh 8 at 1-1.
: 5. The record shows disputed issues of fact in two cate-gories:              a) radiological site suitability matters; and b)
 
                                              - F          environmental considerations.                The Findings which follow are arranged to address the matters in each of these two cate-gories in sequence.
RADIOLOGICAL' SITE SUITABILITY TSSUES
: 6. There are two sets of contested factual issues in regard to radiological site suitability: a) portions of Intervenors' contentions 1,              2, and 3 relating to the site suitability evaluation for severe accidents; and b) Intervenors' conten-tions 2e) and 11d)1) and 2) relating to the validity of the suit suitability dose guideline values recommended by the NRC Staff for CRBRP.              The Findings which follow address these two sets of issues and conclude with Findings as to the ultimate suitability of the CRBRP site.
Contentions 1,            2, and'3 (Site Suitability)
: 7. As admitted by Board rulings, Intervenors' contentions 1,                2, and 3 allege the following: /
: 1.              The envelope of DBAs should include the CDA.
: a. Neither Applicar.ts nor Staf f have demonstrated through reliable data that the probability of anticipated transients without scram or other CDA l
  ~
  */    The Board's Order, dated April 22, 1982, ruled on the scope of inquiry and the applicability (LWA vs. CP) of Inter-venors' contentions 1, 2, and 3.
 
                                - F l 4
initiators is sufficiently low to enable CDAs to be excluded from the envelope of DBAs.,*/
: 2. The analyses of CDAs and their consequences by Appli-cants and Staff are inadequate for purposes of licens-ing the CRBR, performing the NEPA cost / benefit analysis, or demonstrating that the radiological source term for CRBRP would result in potential hazards not exceeded by those from any accident considered credible, as required by 10 CFR I 100.ll(a), n. 1.
: a. The radiological source term analysis used in CRBRP site suitability should be derived through a mechanistic analysis. Neither Applicants nor Staff have based the radiological source term on such an analysis,
: b. The radiological source term analysis should be based on the assumption that CDAs (failure to scram with substantial core disruption) are credible accidents within the DBA envelope, should place an upper bound on the explosive potential of a CDA, and should then derive a conservative esti-mate of the fission product release from such an accident. Neither Applicants nor Staff have performed such an analysis.
  -*/ As admitted by the Board's Order of April 22, 1982, the inquiry at the LWA stage is limited to whether it is feasi-ble to design CRBRP to make hypothetical core disruptive
;    accidents (HCDA's) sufficiently unlikely that they can be
!    excluded from the envelope of design basis accidents for a reactor of the general size and type proposed, upon consid-eration of the following:
: 1. The major classes of accident initiators poten-tially leading to HCDA's;
: 2. The rulevant criteria to be imposed for the CRBRP;
: 3. The state of technology as it relates to applica-ble design characteristics or criteria; and
: 4. The general characteristics of the CRBRP design (e.g., redundant, diverse shutdown systems).
A full scale inquiry into the specific detailed design of
:    the CRBRP is inappropriate at the LWA-1 stage.
l l
l
 
                                              - F                        c. The radiological source term analysis has not ade-quately considered either the release of fission products and core materials, e.g., halogens, iodine and plutonium, or the environmental condi-tions in the reactor containment building created by the release of substantial quantities of sodium. Neither Applicants nor Staff have estab-lished the maximum credible sodium release follow-ing a CDA or included the environmental conditions caused by such a sodium release as part of the radiological source term pathway analysis.*/
i
: d. Neither Applicants nor Staff have demonstrated that the design of the containment is adequate to reduce calculated offsite doses to an acceptable level.**/
: e. As set forth in Contention 8[ll](d), neither Applicants nor Staff have adequately calculated the guideline values for radiation doses from postulated CRBRP releases.***/
i                      f. Applicants have not e'stablished that the computer models (including computer codes) referenced in the Applicants' CDA safety analysis reports, including the PSAR, and referenced in the Staff's CDA safety analyses are valid.            The models and computer codes used in the PSAR and the Staff
!                            safety analyses of CDAs and their consequences have not been adequately documented, verified or validated by comparison with applicable experi-mental data. Applicants' and Staff's safety analyses do not establish that the models accu-rately represent the physical phenomena and
          -*/    Under the Board's April 22, 1982 Order, the limitations set forth for contention la) apply to subparts a)-c) of contention 2. The evidentiary record is to be confined to considering whether the Staff's source term is likely to envelop the design basis accidents as defined under la) for a reactor of the general size and type proposed.
            **/  Under the Board's April 22, 1982 Order, the limitations set forth for contention la) apply to subpart d) of contention 2.
            ***/ Contention 2e), which incorporates contentions 11 d)l) and 2), is addressed separately below.
 
                                        - F                        principles which control the response of CRBR to CDAs.
: g.      Neither Applicants nor Staff have established that the input data and assumptions for the computer models and codes are adequately documented or verified.
: h.      Since neither Applicants nor Staff have estab-lished that the models, computer codes, input data and assumptions are adequately documented, veri-fled and validated, they have also been unable to establish the energetics of a CDA and thus, have also not established the adequacy of the contain-ment of the source term for post accident radio-logical analysis.
: 3. Neither Applicants nor Staff have given sufficient attention to CRBR accidents other than the DBAs for the following reasons:
: b.      Neither Applicants' nor Staff's analyses of poten-tial accident initiators, sequences and events are sufficiently comprehensive to assure that analysis of the DBAs will envelop the entire spectrum of credible accident initiators, sequences and events.
: c.      Accidents associated with core meltthrough follow-ing loss of core geometry and sodium-concrete interactions have not been adequately analyzed.
: d.      Neither Applicants nor Staff have adequately identified and analyzed ths ways in which human l                      error can initiate, exacerbate, or interfere with the mitigation of CRBR accidents.,*/
: 8. Three basic contested issues are presented in regard to site j
suitability by Intervenors' contentions 1,      2, and 3:**/ a)
    --*/  Under the Board's April 22, 1982 Order, the limitations set forth for contention la) apply to subparts b)-d) of contention 3.
      **/  Intervenors have accepted this statement of the issues for these proceedings. TR 6620-21. Accommodation of core melt and disruptive accidents and the environmental effects of (Continued)
 
                                      - F          whether hypothetical core disruptive accidents (HCDA's) should be considered as design basis accidents (DBA's) for CRBRP site suitability analysis (see Intervenors' conten-tions 1 a); 2 a) and b); 3 b) and d)); b) whether the site suitability source term (SSST) for CRBRP results in radio-logical consequences which envelop the spectrum of DBA's (see 10 C.F.R. I 100.11, n. 1; see Intervenors' contentions 2 a), b) (in part) c) and d)); and c) whether the containment design ror CRBRP will reduce of fsite doses to levels within the dose guideline values recommended by the NRC Staff for CRBRP site suitability analysis (see 10 C.F.R. { 100.11; 10 C.F.R. { 50.10 e)2); see Intervenors'                                contention 2 d)) .- /
: 9.        On the affirmative side, the primary sources of evidence addressing these three basic contested issues consist of the NRC Staff's Site Suitability Report (SSR) (S Exb 1), the NRC Staff's Direct Testimony Concerning Site Suitability Issues (S Exh's 2 and 3), the Applicants' Direct Testimony Concern-ing Intervenors' contentions 1,              2, and 3 (A Exh's 1 and 46),
and pertinent portions of the Applicants' Preliminary Safety those events are also encompassed within Intervenors' contentions 1, 2, and 3. The discussion here is confined to the site suitability aspects of those contentions, while the environmental effects of core melt and disruptive accidents are addressed separately below in the Environmental Findings concerning Intervenors' contentions 2d), f), g), and h), and 3 c) and d) (environmental effects of severe accidents).
*/        The validity of the dose guideline values for CRBRP site suitability analysis is addressed below in connection with Intervenors' Contentions 2 e) and 11 d)l) and d)2).
 
                                  - F      Analysis Report (PSAR) and related exhibits (A Exh's 2-19 and 24). On the opposing side, the primary sources of evi-dance consist of Intervenors' testimony by Dr. Cochran (I Exh's 3 and 4) and Dr. Morgan (I Exh 9). The affirmative and opposing evidence differ in their approach to the con-tested issues, particularly in regard to the question of whether the HCDA should be a DBA. The affirmative evidence addresses the contested issues directly in terms of the major classes of initiators and sequences, the general design features, the criteria, and the state of technology for CRBRP, as contemplated by the Board's April 22, 1982 Order, while the opposing evidence addresses the issues in terms of alternative arguments which do not join issue directly with the affirmative evidence. For this reason, the contested issues and the Board's Findings as to each will be first addressed below in the context of the affirma-tive evidence. Following that, the matters raised by the opposing evidence will be specifically addressed.
Whether The HCDA Should Be A DBA.
: 10. The proposed design approach to consideration of accidents in CRBRP is similar in most respects to that normally applied in the Commission's reactor licensing process.
Three levels of safety, incorporating the classical defense-in-depth concept, have been defined (A Exh 1 at 11-12, TR 2000-01; S Exh 5; see 10 C.F.R. Part 50, Appendix A; S Exh 1
 
                                    - F :
l at II-l and Appendix A), and the Staff has established the objective that CRBRP will achieve a level of safety compara-i ble to that of Light Water Reactors (LWR's) (S Exh 1 at II-2: S Exh 5). In addition, majoremphasishasbeenplacep upon the prevention of accidents which could lead to core
!          melt and disruption and loss of containment integrity.      To this end, features and characteristics are incorporated in the design to assure that the likelihood of core disruptive accident initiation is very low, and that the identified accident sequences are terminated within acceptable (design basis) limits -- this is, before they progress to core melt and disruptive conditions. S Exh 1 at II-3; A Exh 1 at 6-8, 11-47, TR 1995-97, 2000-36. On this basis, the Applicants and Staff have proposed that core melting and disruptive accidents can be excluded from the CRBRP design basis.      A q
Exh 1 at 6-8,  11-47, TR 1995-97, 2000-36; S Exh 1 at II-6.
l
: 11. Notwithstanding the forogoing, the CRBRP design approach is l          unique insofar as there are additional provisions in the design to assure that there is a low likelihood of early containment failure and unacceptable consequences associated with core melting and disruptive accidents beyond the design basis. A Exh 1 at 12, 53-73, TR 2001, 2042-62; S Exh 1 at II-16, II-18-19. These provisions include a vent / purge and cleanup system to relieve containment pressure and to limit hydrogen concentrations, a containment annulus cooling I
 
l l
l
_ y_9 _
system, pressure relief vents for the reactor cavity, and associated instrumentation.          S Exh 1 at II-19; A Exh 1 at 55, TR 2044; I. Exh 17.        In addition, the reactor coolant boundary has been strengthened and seals have been added to the reactor vessel closure head to accommodate loadings from core disruption and limit the leakage of sodium, gases and vapors to the containment.          A Exh 1 at 55, TR 2044; A Exh
: 16. These provisions have been evaluated against core disruptive accident conditions, including all pertinent nuclear, thermal, structural and radiological considera-tions, to ensure that the residual risks of such accidents in CRBRP are acceptably low and made comparable to those in
.          LWR's. S Exh 1 at II-18; A Exh 1 at 53-73, TR 2042-62; A 1
1 Exh 46 at 34-35, TR 5410-11.          The proposed approach thus reflects a full consideration of the range of potential accidents, with special emphasis on prevention of core                  1 i
i disruptive accidents and assurance that the residual risk l
associated with such accidents is acceptably low.            In this context, we proceed to consider the question of whether an HCDA should be a DBA.
: 12. Consideration of whether an HCDA should be a DBA starts with examination of the accident initiators and sequences poten-tially leading to HCDA conditions for a reactor of the general size and type as CRBRP.          Once the pertinent initi-l          ators and sequences are identified, one can then assess the
 
                                - F      relevant criteria, the state of technology as it relates to applicable design characteristics or criteria, and the general design characteristics of the CRBRP, and arrive at a judgment as to whether the sequences can be terminated within acceptable limits and before progression to HCDA conditions. See Board Order dated April 22, 1982.      In order i
to provide an initial identification of the relevant acci-dent initiators and sequences, the Applicants conducted an extensive search and developed an accident list based upon relevant previous experience with LWR's, the Fast Flux Test Facility (FFTF), and the NRC Staff's Standard Format and Content of Safety Analysis Reports for Liquid Metal Fast
                        */
Breeder Reactors.-    A Exh 1 at 11-14, TR 2000-3.
: 13. Using the accident list as a point of reference, the Appli-l cants proceeded to a more fundamental physical level of l
analysis. At this level of analysis it was shown, and it was not disputed, that all initiators and sequences of importance to initiation of HCDA conditions - irrespective of their individual causes or details - must involve one or both of two conditions in the reactor core:    a) reduced heat
                                                                        )
i
  -*/ In reaching its judgments, the Staff also relied upon the more than 50 man-years of LMFBR experience on the part of its reviewers, and specific LMFBR operating experience from Fermi, EBR-I and II, SEFOR and FFTF. S W Morris, TR 2394; S Exh 2 at 13, TR 2458.
 
                                                  - F-ll -                                  :
i removal,  or b) excessive heat generation. /  A Exh 1 at 14-15, TR 2003-4, A W Clare TR 1836, 1852; S Exh 1 at II-6; S Exh 2 at 14-15, TR 2459-60. This analysis examined the full spectrum of potential sequences which could evolve from these two reactor core conditions and the general design characteristics of CRBRP, and showed, without dispute, that four systems or sets of design features are necessary to maintain core conditions so that reduced heat removal or excessive heat generation will remain within design basis conditions and not progress to HCDA conditions. S Exh 1 at II II-13; S Fxh 2 at 10-11, 13-19, TR 2455-56, 2458- 64;
                  -      A Exh 1 at 14-46, TR 2003-2035. No evidence is present in the record to suggest that the pertinent initiators and sequences have not been adequately identified and considered, nor is there any evidence of any additional credible scenarios which were not considered.
: 14. The four systems or sets of features of undisputed importance to whether an HCDA should be a DBA are:    a) the Reactor Shutdown Systems (RSS); b) the Shutdown Heat Removal Systems
                  --*/  The Staff's approach was substantively the same, and differed only as a matter of convention. The Staff defined four basic conditions or failures of importance to HCDA initiation: a) failure to shut down the nuclear chain reaction during an overpower or flow reduction transient; b) failure to maintain sufficient coolant inventory; c) failure to maintain suffi-cient coolant flow; and d) failure to extract sufficient heat from coolant. S Exh 2 at 10-11, TR 2455-56. Failure a) involves excessive heat generation or reduced heat removal, while failures b), c), and d) involve reduced heat removal.
i
 
                                - F l (SHRS); c) means to prevent double-ended rupture of Primary Heat Transport System (PHTS) inlet piping; and d) features to prevent local imbalance between heat generation and heat removal from progressing to significant core-wide involve-ment. Id. The general design characteristics for each of these four systems or sets of features were evaluated by the NRC Staff and Applicants in light of the relevant criteria for CRBRP and the state of technology (as it relates to applicable design characteristics or criteria), in arriving at their judgments that HCDA's are sufficiently unlikely that they can be excluded as DBA's for the purposes of CRBRP site suitability analysis. Id. The Board accepts those judgments as reasonable and as clearly supported by the weight of the evidence in the record. Id. The record evidence which shows that each of these systems or sets of features can be relied upon to exclude HCDA's as DBA's is the subject of the Findings which follow.
Reactor Shutdown' Systems
: 15. Excessive heat generation in CRBRP requires a reactivity insertion to the reactor. A Exh 1 at 20-25, TR 2009-14; A Exh's 9 and 10; A Exh 46 at 5, TR 5381. In the reactivity regime of importance to prevention of excessive heat genera-tion leading to HCDA conditions, the reactor kinetics and l      reactivity control characteristics of LMFBR's, including CRB RP , are essentially similar to those of LWR's. A Exh 46 l
 
                                        - F      at 6-13, TR 5382-89.            Reduced heat removal in CRBRP could result from reduced primary coolant flow through the reactor-core and/or increased primary coolant temperature at the core inlet.            A Exh 1 at 15-20, TR 2004-2009; A Exh 11; A Exh 46 at 13, TR 5389.            To assure a high likelihood that HCDA's caused by an imbalance between heat generation and heat removal can be prevented, CRBRP has proposed a design consisting of two (rather than one, as in LWR's) fast-acting reactor shutdown systems (the primary and secondary shutdown systems), either of which can reduce reactor power level (and hence restore the balance between heat generation and heat removal) and shut down the reactor when required.-l            S Exh 1 at II-7, A Exh 1 at 27-35, TR 2016-24; A Exh's 3 and
: 10. The design of the systems will conform to all applicable
  -*/ Intervenors have advanced only one apparently substantive argument in opposition to the Applicants' and Staff's judg-ments concerning the RSS.            Intervenors point to the Commis-l sion's proposed anticipated transients without scram (ATWS) rule (46 Fed. Reg. 57521 (November 24, 1981)) for the proposition that existing LWR reactor protective systems are not adequate, and by implication, that the CRBRP systems are not.            I Exh 3 at 36-37, TR 2845-46. This proposition cannot withstand scrutiny since: a) the Commission has found that, although reductions should be made in the severity and fre-quency of ATWS events in LWR's, the likelihood of severe consequences pending implementation of the rule is small and there is reasonable assurance of safety pending implementa-tion of the rule (46 Fed. Reg. 57522); and b) CRBRP will have two rather than one RSS as in LWR's. (S Exh 1 at II-7). A 10-3 ATWS frequency would not be applicable to CRBRP.            S W Rumble, TR 2417; S W Morris, TR 5450.
Intervenors' Witness Cochran had never even reviewed an LWR RSS.            I W Cochran, TR 2664.
 
                                                                            - F ,
criteria in the CRBRP Design Criteria, NRC Regulatory Guides l
;                                                  and IEEE Standards. S Exh 1 at II-7; S Exh 1, Appendix A; see 10 C.F.R. Part 50, Appendix A. The two shutdown systems are automatically actuated, fast-acting, and redundant, diverse, and independent in regard to sensors, logic, control i                                                  rod drive mechanisms, and control rods.          S Exh 1 at II-6-II-8; A Exh 1 at 27-35, TR 2016-2024; A Exh's 3 and 10; S Exh 2 at 15-16, TR 2460-61. The systems are fail-safe -- no elec-trical or cxternal power is required for a scram of any control rod, and loss of power will initiate a scram.                Id.
The Applicants have proposed, and the NRC Staff will impose, a requirement for a hardware-oriented, confirmatory relia-bility program to assure that the potential for high relia-bility embodied in these design characteristics and criteria i                                                  is actually realized for CRBRP. S W Morris, TR 5646-47; S
!                                                Exh 8 at 12-78; S W's Morris and Rumble, TR 5518-21.              Given these general design characteristics, criteria, and the cur-rent state of technology, it is clearly feasible to imple-ment a reliable RSS as proposed for CRBRP which will assure that initiation of HCDA conditions through an imbalance between heat generation and heat removal is highly unlikely.
Shutdown Heat Removal Systems
: 16.                    The shutdown (or residual) heat removal system (SHRS) design proposed for CRBRP will remove reactor core decay heat (and hence prevent reduced heat removal) through four separate y-,    . . - _ . - . , . . . _ , . . . . _
_ - - . - _ - - - . - . , . ~ _ , _ _ _ - _,                                        7 _.m- . , -    - . _ _,
 
                                - F        heat transport paths:  a)  through each of three Primary Heat Transport Systems (PHTS) and Intermediate Heat Trans-port Systems (IRTS) loops with the Steam Generator Auxiliary Heat Removal System (SGAHRS); and b) by means of a diverse Direct Heat Removal Service (DHRS). Any one of the four paths is capable of removing all short and long term decay heat from the reactor core. S Exh 1 at II-ll - II-13; A Exh l      1 at 35-40, TR 2024-29; S Exh 2 at 18-19, TR 2463-64; A Exh l
: 4. These systems, which conform to CRBRP Design Criterion 35 and incorporate redundant, diverse, and independent features, are safety grade systems. Id.; S Exh 1, App. A;
;      see 10 C.F.R. Part 50, Appendix A, General Design Criterion
: 34. The three heat removal paths provided by the SGAHRS system use well proven technology similar to that of
              */
LWR's.-  A Exh 1 at 35-38, TR 2024-27; A W Brown, TR 1726-29; A W Clare, TR 1834-35, 5269; S W Rumble, TR 5559-60.
i Although the DHRS has no direct counterpart in LWR's, the I
proposed use of well-proven components in this system is      -
i within the state of the art. A Exh 1 at 38-39, TR 2027-28;  !
A W Brown, TR 1726-29; A W O' Block, TR 1643-46; S W Rumble, TR 5581; A W Clare, TR 5278-79, 5286-87. Although the NRC l
l
  -~*/  The inherent characteristics of the CRBRP indicate that one can expect a higher reliability in this regard, relative to LWR experience. A Exh 46 at 14-20, TR 5390-96.                l
 
                            - F    Staff has not yet completed its in-depth safety review of the SHRS, and has reserved judgment as to the adequacy of its diversity, experimental data from FFTF indicate that natural circulation cooling will provide the necessary diversity. S Exh 1 at II II-13; A W Clare, TR 5278; A W Strawbridge, TR 5001-65. In any event, should the Staff's review conclude that natural circulation cooling will not provide the necessary diversity, the Staff will impose a condition for providing motive and control power throughout the entire SHRS train in the event of loss of offsite and onsite AC power supplies. S Exh 1 at II-13. As with the RSS, the Applicants have proposed, and the Staff will impose, a hardware-oriented, conformatory reliability pro-gram which will assure that the potential reliability embodied in the design and criteria will actually be real-ized for CRBRP. S W Morris, TR 5646-47; S Exh 8 at 12-78; A W Clare, TR 1689-90. Given these general design charac-teristics, criteria, and the = tate of technology, it is clearly feasible to implement a reliable SHRS as proposed for CRBRP which will assure that initiation of HCDA conditions due to failure to remove decay heat is highly l
unlikely.
 
                              =
l l
                                    - F      Means' to'' Prevent Double-Ended' Rupture of' Primary Heat' Transport System Inlet Piping l
: 17. The proposed CRBRP design has incorporated a series of measures to prevent a double-ended pipe rupture in the PHTS inlet piping, a resultant leak in excess of design basis values, and sodium boiling in the core (and hence reduced heat removal). S Exh 1 at II-8-II-9; A Exh 1 at 40-43, TR 2029-32. Since the operating temperature of the sodium coolant (700-1000* F) is below the saturation temperature (1600* F), the sodium coolant is pressurized only to the extent necessary to pump the coolant through the primary l
system, and there is no potential for flashing of the coolant into vapor due to loss of system pressure as in an LWR. S Exh 1 at II-8; A Exh 1 at 40-43, TR 2029-32; A Exh 46 at 14-16, TR 5390-92; S Exh 2 at 16-17, TR 2461-62.                        The
;          CRBRP primary coolant system piping and components are housed in a series of inerted cells, which provides protec-tion and a less hostile environment (from the standpoint of degradation processes) than in LWR service.            S Exh 1 at II-8; A Exh 1 at 40-41, TR 2029-30; A Exh 46 at 15-16, TR 5391-92; S Exh 2 at 16-17, TR 2461-62. Nevertheless, measures to prevent rupture of the cold-leg piping within a section near the reactor vessel inlet have been provided in the design. /
      --*/  Rupture of cold leg piping could result in a flow reversal in the core and sodium boiling (with consequent reduced heat removal), and in that event, could lead to an HCDA. See A (Continued) l
 
l
                          - F l The CRBRP design characteristics incorporate a series of measures to assure that very small pipe leaks can be detected and that inlet piping leaks in excess of design basis values can be prevented. These measures include the use of tough ductile stainless steel, proven for high temperature service, as a piping material, and redundant, diverse leak detection systems which can detect a leak several orders of magnitude below design basis values.        A Exh 1 at 40, TR 2029; S Exh 1 at II II-9; S Exh 2 at 16-17, TR 2461-62. In addition, the CRBRP has:        a) specified stringent limitations on flaws in the piping materials to assure that the potential for crack initiation is minimized; b) conducted fracture mechanics analyses to show that, if a flaw up to the material specification limits existed, it would not propagate to create a crack in the piping wall; c) completed a technology program to show that even if a crack develops, it can be detected, and d) conducted additional i
l analyses to show that any crack which would develop would have dimensions well below that analyzed for the design Exh 1 at 40, TR 2029; S Exh 1 at II II-9.        Up to this point in its review, the Staff has treated a hot leg pipe rupture as a design basis event. S Exh 1 at II-9. The consequences of this event, however, do not result in sodium boiling and reduced heat removal to the extent that an HCDA could be initiated. Therefore, this event would have sub-stantially less importance to the question of whether an HCDA should be a DBA. S Exh 1 at II-9; A Exh 11; S Exh 2 at 17, TR 2462.
l l
                                              ,-----r-
 
                                    - F        basis leak size. A Exh 1 at 40-43, TR 2029-32; S Exh 1 at II-9; S Exh 2 at 16-17, TR 2461-62; A Exh 24.      These measures will be complemented by a pre-sorvice and in-service inspection program, a material surveillance program, and verification of leak detection system performance.        S Exh 1 at II-9. Given these design characteristics, cri-teria, and the state of technology, it is clearly feasible to provide primary piping integrity as proposed for CRBRP which will assure that initiation of HCDA conditions through reduced heat removal is highly unlikely.
Features and Capabilities to Prevent Local Imbalance Between Heat Generation and Heat Removal
: 18. The CRBRP design incorporates features and inherent capa-bilities to assure that local imbalances between heat generation and heat removal do not progress to involvement of a significant portion of the core (and hence HCDA initi-ating conditions).      S Exh 1 at II II-ll; A Exh 1 at 43-46, TR 2032-35; S Exh 2 at 19-20, TR 2464-65.      These features and capabilities are of two types:      a) those to preclude a rapid reduction in flow to a limited region of the core; and b) those to ensure that local fuel pin i        failures would not propagate to involve a significant I
portion of the core.      A Exh 1 at 43, TR 2032. Features of the former type consist of a multiplicity of flow paths in the fuel subassembly inlet, in the inlet modules that hold 1
l
 
l
                                      - F              groups of subassemblies, and in the core support structure that holds and supports the inlet modules. These redundant flow paths provide an inherently reliable, passive means of
!            preventing flow reduction to a fuel assembly caused by foreign objects. A Exh 1 at 44-45, TR 2033-34; S Exh 1 at II-10; S Exh 2 at 18-19, TR 2462-63. These features, complemented by a detection capability provided by core outlet thermocouples, would prevent a large sudden flow reduction such as that which led to damage to two fuel subassemblies in the Fermi reactor. /    S Exh 1 at II-10; A W Strawbridge, TR 1828-30; A Exh 12. Features and capabili-ties of the latter type (those to preclude propagation of local failures) rest upon a broad base of experimental and analytical work at EBR-II, world-wide LMFBR operating experience, and analyses of the CRBRP design cl.aracteris-tics, all of which show a very low likelihood that local
,            fuel rod failures would propagate beyond their immediate i
vicinity. A Exh 1 at 45, TR 2034; S Exh 1 at II-10-II-ll; A Exh 12; S Exh 2 at 19-20, TR 2464-65; S Exh 17 at 9-11, TR 5756-58. In addition, steel hexagonal subassembly ducts V
  */
  -~
The CRBRP design contemplates filtration of particulates during startup testing and maintenance of purity by the cold trap purification system. A Exh 1 at 44-45, TR 2033-34.
Nevertheless, the Applicants have assessed the consequences of blockages and have shown that with more than 80% of the flow area blocked, sodium boiling and HCDA initiation would not result. A Exh 1 at 45, TR 2034; A Exh 12.
I    - _ .
 
                                              - F                house each fuel rod bundle and provide passive, inherent physical protection against propagation from one subassembly to another. A Exh 1 at 45, TR 2034.      Finally, fuel failures can be detected by fission gas detectors and delayed neutron detectors at levels below those that could result in local blockages. Id. The features and capabilities to prevent local imbalances will be complemented by the use of core outlet thermocouples for detection, an operational fuel surveillance program, and, if necessary, monitoring and operating restrictions, to assure that fuel failure propaga-tion will be prevented by the design and operating charac-teristics. S Exh 1 at II-ll.        Given these design charac-teristics, criteria, and the state of technology, it is clearly feasible to maintain local heat generation / heat i              removal balances as proposed for CRBRP and assure that initiation of HCDA conditions from local failures is highly unlikely.
: 19. The criteria applied in consideration of whether HCDA's should be DBA's in CRBRP rest 'pon two sound basic prin-ciples:  a) the design approach should incorporate the well established defense-in-depth concept ( e . g . , A E xh 8 ) , and b) the risks associated with accidents in CRBRP (including beyond design basis accidents) should be comparable to those for LWR's (e.g., S Exh 1 at II-2).        In applying the first of
 
                              - F    these principles, the Applicants have established require-ments for redundancy, diversity, and independence of systems for prevention of HCDA initiation that go beyond the equiva-lent requirements for LWR's. A Exh 1 at 26-46, TR 2015-35; A Exh 46 at 12, 16, 19, TR 5388, 5392, 5395.      In applying
!  the second of these basic principles, the Applicants have established requirements for design features aimed at limit-ing the residual risks of HCDA's. A Exh 1 at 54-55, TR 2043-44. The Staff's {{letter dated|date=May 6, 1976|text=May 6, 1976 letter}} establishes a set of general requirements for the design safety approach, site suitability source term and core disruptive, accidents which have been implemented in the Staff's review.      S Exh 5. The Staff's evaluation developed a set of criteria which are appropriate in light of the general design characteristics of CRBRP; namely, to assure that the design will incorporate a) redundant, independent, and diverse reactivity shutdown systems; b) redundant, independent, and diverse heat removal systems; c) means to detect and prevent propagation of local fuel faults; and d) assurance of continuing high integrity of the heat transport system. S Exh 2 at 14. TR 2459. The Staff will apply a set of CRBRP Design Criteria which are similar to those in 10 C.F.R. Part 50, Appendix A (with j  modifications to reflect technological differences).      S Exh l
1, Appendix A; see 10 C.F.R. Part 50, Appendix A.      These criteria impose requirements for redundancy, diversity and
 
                              - F      independence in the pertinent CRBRP safety features, which will be complemented by the Staff's additional requirement for a reliability program to assure that the reliability inherent in these systems and features is realized.      SW Morris, TR 5646-7; S W's Morris and Rumble, TR 5518-21; S l
Exh 8 at 12-78. These general criteria will be coupled with the specific criteria embodied in the existing generic NRC regulatory requirements (e.g.,    10 C.F.R. Part 100, 10 C.F.R.
                                                        */
Part 73, 10 C.F.R. Part 50, Appendices B and R)- and incorporated in the Staff's ongoing safety review.      S Exh 2 at 20-23, TR 2465-68. Altogether, the body of requirements described above represents an adequate and sufficient set of criteria for use in the determination as to whether an HCDA should be a DBA for purposes of CRBRP site suitability analysis.
: 20. The potential for, and actions to minimize, human error and common cause failures have been considered and implemented in the design to assure that the likelihood that system i
interdependencies or human error could cause an HCDA is made extremely low. S W Morris, TR 2256; S Exh 2 at 15-16, 20-23, 23-25, TR 2460-61, 2465-68, 2468-70; S Exh 8 at 12      12-78; A Exh 1 at 13-14, 2002-03. The general design characteristics of CRBRP include the use of:    a) redundant,
  */  S Exh 2 at 20-23, TR 2465-68.
I
 
                                  - F          independent, and diverse and automatically actuated or pas-sive safety systems, and b) inherent physical characteris-tics which assure that rapid operator action will not be necessary in responding to accidents at CRBRP, and that the potential for human error will be minimized. S Exh 2 at 15-25, TR 2460-70; A Exh 1 at 14-46, TR 2003-35; A Exh 46 at 5-22, 29-33; TR 5381-98, 5405-U9; S W Morris, TR 2221-25. The Staff's review of the design will account for system inter-dependencies and common cause failures by reliance upon well established principles enunciated in IEEE Standard 279 and the applicable Standard Review Plan. S Exh 2 at 15-16, TR 2460-61. The Applicants have undertaken an extensive series of systems interaction studies, such as key systems reviews, to assure that human error, system interdependencies and common cause failures will not compromise the reliability inherent in the redundant, diverse and independent systems of importance to the prevention of HCDA's. A W Clare, TR 5247-49. The Applicants have proposed, and the Staff will require, implementation of a reliability program to assure that the reliability inherent in the CRBRP design char-acteristics will be realized and not degraded by potential common cause failures. S W Morris, TR 5646-47; S Exh 8 at 12-77-12-78. The record shows uncontroverted evidence that human error and common cause failures have been and will be extensively considered by the Applicants and NRC Staff and
 
                                    - F l that neither is likely to affect the present judgment that HCDA's need not be considered as DBA's for purposes of CRBRP site suitability analysis.
: 21. On the basis of the foregoing Findings 10-20 and the evidence in the record as a whole the Board finds that:  a) the sequences and conditions of importance to HCDA initiation have been properly identified; b) the envelope of design      '
                                                                    )
basis accident conditions and events has been properly iden-  ,
I tified and described; c) the systems and design features of importance to maintenance of core conditions within the design basis and prevention of conditions for HCDA initia-    ,
l' tion have been properly identified and described; d) those systems and features incorporate the appropriate design        ,
I characteristics, criteria, and technology for prevention of HCDA's; e) in light of the design characteristics, criteria,  ,
l and state of technology, it is highly unlikely that HCDA      l l
conditions will occur in CRBRP; f) it is feasible to design    ;
i and implement systems and features such that HCDA's need not be considered DBA's in CRBRP; g) human error and common        l cause failures have been considered and neither is likely to affect the judgment that HCDA's need not be considered as DBA's; and h) HCDA's are not credible events and need not be considered DBA's for the purposes of CRBRP site suitability analysis.
                                                                    )
I
 
                                - F l
: 22. While the likelihood of progression to HCDA conditions is very low, the residual risk associated with such conditions is also very low. Thre pertinent nuclear, thermal, hydraulic, structural, and radiological considerations attending HCDA's are well understood and have been evaluated using proven state-of-the-art techniques, the,v$lidityof which stands uncontroverted in the record.                                    A Exh 1 at 53-73, TR 2042-62; A Exh 46 at 34-35, TR 5410-11.                                    The most likely progression of an HCDA results in a non-energetic termination, and substantially pessimistic assumptions must be invoked to predict significant energetics.                                    A Erh 1 at 56-63, TR 2045-52; A Exh 15.      Further, analyses of the l
provisions in the design for accommodation of energetic disruption indicate that a design meeting the structural criteria for accommodation of HCDA's can bN implemented.                                    A Exh 1 at 63-65, TR 2052-54; A Exh 16.                                Even if core damage is presumed to progress to core melting and thermal penetration of the reactor vessel and guard vessel, the analyses (which include adequate consideration of sodium-concrete reactions ( A Exh 1 at 67, TR 2056; A Exh 19))
nevertheless show that containment integrity would be maintained for about a day following the.' initiation of core melting (assuming multiple failure of design basis systems and design features, and no action to mitigate the event).
i l
 
                                  - F        A Exh 1 at 65-67, TR 2054-56; A Exh 18.      Moreover, contain-ment integrity can be maintained indefinitely by controlled venting and purging through a cleanup system and by cooling the annulus between the containment and confinement building using the provisions in the design for accommodation of core melt. Id.; A Exh 17. The radiological consequences of l
HCDA's have been analyzed over a broad range of conditions and sequences, and these analyses show that radiological consequences are acceptably low, are not very sensitive to f?  *4 me of venting or the amount of release through the
        .  .cor head, and are characterized by atmospheric releases which are comparable to those for similar core melt sequences in LWR's. A Exh 1 at 67-73, TR 2056-62; A Exh 46 at 32-36, TR 5408-12.      Our Finding here, that the residual risk from accidents beyond the design basis can be made acceptably low and comparable to LWR's, lends substantial confidence to our findings that HCDA's need not be consid-ered as DBA's for purposes of CRBRP site suitability analysis. A Exh 1 at 53-73, TR 2042-62; S Exh 8, Appendix J; See Findings 10-21, supra.
l The Site Suitability Source Term Results In Consequences Which~ Exceed Those~For DBA's
: 23. The NRC Staff selected a site suitability source term (SSST) which is directly analagous to that for LWR's. S Exh 1 at III-8. The SSST consists of the usual LWR source term to be
 
                                                                                            - F                                        released from the core, plus 1% of the plutonium in the core.          S Exh 1 at III-8; A Exh 1 at 48-49, TR 2037-38; S Exh 3 at 8-10, TR 2491-93; see 10 C.F.R. $ 100.ll(a), n. 1; S Exh 3, Attachment A.                                The SSST prescribes an inventory consisting of 100% of the noble gases, 50% of the halogens (half of the iodine isotopes are assumed to plate out within a short time period), 1% of the solid fission products, and 1% of the plutonium,-/ which is assumed to be mixed uniformly throughout the primary containment and instantaneously available for release to the environment through containment structures, filters, and engineered safety features.                                      Id.;
see S Exh 1 at III III-9.                                      The SSST for the CRBRP is a non-mechanistic term which represents an assumed release from the core whose consequences (based on the uncontra-dicted evidence in the record) envelop and are not exceeded by any accident considered credible; i.e., any release from the core within the DBA envelope. S Exh 1 at III-9; S Exh 3 at 8-15, TR 2491-98; A Exh 1 at 47-48, TR 2036-37; see 10
*/                                            Intervenors have argued that the plutonium source term should be 10%, I Exh 4 at 22, TR 3072. Values of this magnitude, however, have been postulated only for extremely conservative analyses of energetic HCDA's. Id.; see, A Exh 1 at 53-73, TR 2042-62.                                  The 1% value selected by the NRC Staff is appropriately conservative (S Exh 3 at 13-15, TR 2496-98), and in view of our Findings 10-21, supra, the 10%
value urged by Intervenors is not warranted.
 
                                                      - F                      C.F.R. $ 100.ll(a), n.1; see Findings 10-21, supra.                Accord-ingly, and in light of Findings 2-21, supra, the conse-quences resulting from the SSST release envelop those asso-I ciated with DBA's for the purposes of CRBRP site suitability analysis.                                                                  ,
1 i
l The CRBRP Containment Can Limit SSST Doses Within                                      !
The Site Suitability Dose Guideline Values
: 24.      Given the SSST, CRBRP site meteorological data in accordance with NRC Regulatory Guide 1.145, and a design basis leak rate for the containment of 0.1 volume percent per day, the calculations performed by the NRC Staff and Applicants show that the containment can limit offsite doses within the site suitability dose guideline values.-/              S Exh 1 at II-ll; S Exh 3 e.t 15, TR 2498; A Exh 1 at 51,,TR 2040; S W Bell, TR
!                    2418; A W Strawbridge, TR 1838; A Exh 14.              The computer models used by the NRC Staff and Applicants for SSST                      ,
1 analysis, including their input data and assumptions, have 1
i
        --*/        The effects of sodium were properly considered.                The assump-tion of zero sodium in connection with the SSST release is conservative since the presence of sodium would increase the rate of agglomeration and fallout of aerosols, reduce the amounts of halogens and plutonium available for release, and result in lower offsite doses. A Exh 1 at 49, TR 2038; S Fxh 17 at 29, TR 5777. The maximum sodium release would result in an increase of containment pressure up to 2 psig                ,
(versus 10 psig design pressure), while the SSST analysis                  :
assumes a design leak rate corresponding to a pressure of 10 psig over the entire 30 day period of release. A Exh 1 at 49-50, TR 2038-39.            The potential dose contribution of sodium radioisotopes in the SSST would be negligible in the context of SSST analysis. S Exh 3 at 17-21, TR 2500-04.
 
F      been adequately documented, verified and validated.                          S Exh 3 at 36-41, TR 2519-24; A Exh 1 at 51-52, TR 2040-41; A Exh
: 14.                There is no reliable, probative evidence in the record to the contrary.                  Ther2 is no substantial evidence in the record to controvert either the validity or the applicabil-ity of the meteorological data.                  S Exh 1 at IV-1-IV-2.        Nor is there any such evidence to controvert the capability of                          )
the CRBRP containment to achieve a 0.1 volume percent per day leak rate, as demonstrated by tests on similar LWR containments and the FFTF.                  A Exh 1 at 50-51, TR 2039-40; S Exh 3 at 22-25, TR 2505-08; A Exh 5; A Exh 14.                    The Board finds that the CRBRP containment can limit SSST doses within the site suitability dose guideline values.
: 25. . Based upon the foregoing Findings 10-24, and subject to those Findings below concerning the specific issues raised by Intervenors' opposing evidence, / the Board finds that Applicants have sustained their burden as to each of the
*/    The areas of dispute consist of Intervenors' arguments
-~
that:                  a) the HCDA should be a DBA; b) the SSST should be predicated upon including the HCDA within the envelope of DBA's; and c) the methods of analysis and assumptions used by the Applicants and Staff for the SSST dose calculations were not valid. The first two arguments have been disposed of by Findings 10-22 supra.                  The remaining specific matters in dispute concerning t:1e validity of the methods of analysis and assumptions for the SSST dose calculations are addressed below in connection with our consideration of the opposing evidence advanced by Intervenors.
 
                                                                                                                                                                                            - F                                                                                                                                                        three basic questions encompassed within Intervenors' con-tantions 1, 2, and 3 (site suitability).              That is, for purposes of CRBRP site suitability analyses under 10 C.F.R.
9 100.11 and 10 C.F.R. I 50.10e)2): a) the HCDA need not be considered a DBA (Findings 10-22); b) the SSST selected by the NRC Staff envelops the consequences of releases from the core as a result of DBA's (Finding 23); and c) the contain-ment can limit SSST doses within the site suitability dose guideline values recommended by the NRC Staf f for CRBRP site suitability analysis (Finding 24).
: 26.                                                                                                      Intervenors' opposing evidence can be categorized in two broad subject areas or classes of arguments:              a) the HCDA should be a DBA; and b) the methods of analyses and assump-tions for the Applicants' and NRC Staff's SSST analyses are invalid.              Within each of these two subject categories, the Intervenors have raised a series of specific issues.              In what follows, the specific issues raised within each of the two subject categories will be addressed and Findings rendered on the basis of the evidence in the record as a whole.
Intervenors' Arguments Relating To Why The HCDA Should Be Considered A DBA For CRBRP Site Suitability Analysis
: 27.                                                                                                  Intervenors have advanced four primary lines of argument in support of their position that HCDA's should be DBA's for CRBRP site suitability analysis.              According to Intervenors,
 
                                                              - F t for the purposes of CRBRP site suitability analyses:                                                a) previous experience with domestic sodium cooled reactors indicates that an HCDA should be a DBA; b) previous experience with foreign sodium cooled reactors indicates that an HCDA should be a DBA; c) previous experience with the CRBRP parallel design, in which the HCDA was a DBA, 4
indicates that the HCDA should be a DBA; and d) quantitative                                                      1 and probabilistic analyses, which have not been relied upon by the NRC Staff or Applicants for this purpose, are necessary for a decision concerning whether an HCDA should be a DBA for purposes of site suitability analysis.- /                                              Each of these arguments is addressed in succession in the following Findings.
Domestic Sodium Cooled Reactor Experience
: 28. Intervenors point to operating and licensing experience with five domestic sodium cooled reactors -- EBR-I, EBR-II, Fermi, SEFOR, and FFTF -- as the basis for arguing that HCDA's should be DBA's for CRBRP.                                    I Exh 3 at 13-19, TR
  -*/  Intervenors have also argued that neither Applicants nor Staff have given adequate consideration to human error or common cause failures.                                  (See I Exh 3 at 40-42, TR 2849-51; TR 6697-98), but they have not advanced'any reliable evidence in support of that proposition. Against this mere assertion of inadequate consideration, the record contains evidence i        that both Applicants and Staff have given extensive consid-eration to these factors, and that neither of these factors is likely to affect the present judgment that HCDA's should not be DBA's. See Finding 20, supra.
 
                                  - F ,
2822-28. In each case cited by Intervenors, however, the experience does not support the proposition that HCDA's should be DBA's for CRBRP site suitability analysis, l  29. EBR-I, an experimental LMFBR that initially operated in 1951, incurred core damage during an experiment in 1955.      I Exh 3 at 13-14, TR 2822-23. Intervenors' witness was not even aware, however, that the shutdown system was discon-nected to permit an intentional power excursion, and that the operator did not respond properly to terminate the excursion with the remaining reactor protection system. I W Cochran, TR 2628-30; S Exh 2 at 26, TR 2471. No significant releases of radioactivity resulted. I W Cochran, TR 2628-30; S Exh 1 at II-3. In light of the circumstances of this incident and the lack of currency and applicability of the experience to CRBRP (S Exh 8 at 12-32), it does not support a finding that HCDA's should be DBA's.
: 30. EBR-II, an experimental, government-owned LMFBR which commenced operations in 1963, included calculations of energetic core disruptive accidents as part of its hazards analysis. I Exh 3 at 14, TR 2823. While the lack of currency of this experience throws this evidence into doubt, the record is also devoid of evidence to show that the general design characteristics, criteria, and state of tech-nology for EBR-II were in any way similar to those for CRBRP wY
 
                              - F                              l
                                              */
and considered in Findings 10-21, supra.-    This experience does no more than show what was done for EBR-II. There is no factual nexus between this experience and CRBRP to        ,
l support a finding that HCDA's should be DBA's for CRBRP.      ,
: 31. At the Fermi reactor an incident occurred in which two fuel I
subassemblies partially melted as a result of blockage of the assemblies by a locse plate within the reactor system.
S Exh 1 at II-3-II-4. Intervenors argued that the design basis for Fermi (melt of one fuel assembly) was exceeded, and that therefore HCDA's should be DBA's for CRBRP. I Exh 3 at 14-15, TR 2824. The incident, however, did not result in any release of radioactivity to the environment. S Exh 1 at II-4. Moreover, the total amount of fuel partially melted in both of the two subassemblies did not exceed the equivalent amount in one subassembly, and the design basis for Fermi does not appear to have been exceeded. IW Cochran, TR 2635-37. The most telling point, however, is that the CRBRP includes design features which incorporate the lessons learned from, and would preclude the recurrence of, an event similar to the Fermi incident. Finding 18,
  */ Intervenor's own direct testimony raises a substantial ques-tion as to whether an energetic HCDA was a DBA for EBR-II.
Intervenors allege that EBR-II performed analyses for an event with energetics of 1050 pounds TNT equivalent, but at the same time the containment was designed to withstand 300 pounds TNT equivalent. I Exh 3 at 14, TR 3823; I W Cochran, TR 2633-34.
 
i
                                        - F I supra; S Exh 1 at II-3-II-4; S Exh 2 at 26,'TR 2471; A W Strawbridge, TR 1828-30.            The Fermi experience does not support a Finding that HCDA's should be DBA's for CRBRP.
: 32. SEFOR, an experimental LMFBR, which initially operated in 1969, considered energetic core disruptive evente as part of its safety analysis.          I Exh 3 at 15-16, TR 2824-25.              While the record does not show that these events were in fact treated as DBA's per se (contra, A W's Brown and Clare, TR 1825-26), the nature of the facility itself indicates that SEFOR design criteria cannot be meaningfully applied to CRBRP. SEFOR was a test reactor specifically designed to undergo rapid power excursions to develop basic physical data on the theretofore untested Doppler coefficient of reactivity.          SW Becker, TR 2396-97; I W Cochran, TR 2638-
: 39. In contrast, CRBRP is designed and intended as a baseload unit, and its reactivity characteristics, including the Doppler coefficient, are well understood and supported by experimental data.          A Exh 46 at 5-10, TR 5381-86; A Exh 58 at 14-15, TR 6420-21.            In light of this, the SEFOR experience is not applicable and does not support a finding that HCDA's should be DBA's.
: 33. FFTF, a government-owned reactor for testing LMFBR fuels, underwent review by the AEC Regulatory Staff, and later the NRC Staff, prior to cc.6mencement of operation in 1980.                      I Exh 3 at 13, TR 2822; S Exh 1 at II-4.                      Intervenors cite
 
                                  - F      numerous documents from that review for the proposition that l
;    HCDA's were DBA's for FFTF, and by implication, should be DBA's for CRBRP.        I Exh 3 at 16-18, TR 2825-27. While the portions of these documents cited and quoted by Intervenors are ambiguous at best, expert witnesses with first-hand knowledge of the FFTF project explaine'd how HCDA's were treated in regard to the FFTF design basis.        In fact, the approach taken for FFTP was in substance the same as that f or CRB RP , and although the approach was articulated some-what differently, in practice, HCDA's were not treated as DBA's.        A W Brown, TR 1501; A W's Brown and Clare, TR 1825-26; S W King, TR 2395; S W Long, TR 2396.        Thus, the most recent and most closely applicable domestic reactor experience does not support a finding that HCDA's should be DBA's for CRBRP.
: 34. Aside from the fact that the evidence of domestic reactor experience relied upon by Intervenors does not support the propocitions asserted, there is also considerable doubt as to its reliability and probative value.        While previous domestic reactor experience can surely be relevant to a decision here, the type of experience advanced by Inter-venors for EBR-I, EBR-II, Fermi, SEFOR, and FFTF, as dis-cussed above, does not represent the kind and quality of reliable, probative avidence upon which the Board can or should base a decision of this moment.
1 l
 
                                                      - F  Foreign Reactor Experience
: 35. Intervenors point to experience with the French Super Phenix and German SNR-300 reactors in arguing that HCDA's should be
,      DBA's for CRBRP.                          I Exh 3 at 19, TR 2828. As for the French Super Phenix reactor, Intervenors' Witness Dr. Cochran asserted that HCDA's were treated as DBA's and that the con-tainment design concept is the equivalent of the previous (superseded) CRBRP parallel design with a sealed head access area and core catcher.                          I Exh 3 at 19, TR 2828. The record shows, however, that the French treat HCDA's as outside the design basis for 5uper Phenix, and that the approach taken is essentially the same as that for CRBRP.                            Compare A Exh 29 with S Exh 5; I W Cochran, TR 2649.                            In addition, the Super Phenix containment design is not the equivalent of the CRBRP parallel design sealed head access area.                            Rather, the Super Phenix design shows a dome-like single shell inner containment with a surrounding concrete confinement struc-ture.      I Exh 5 at 1.                      Dr. Cochran's information about the Super Phenix containment consisted merely of one technical article, the characteristics of which he misconceived (see I Exh 5 at 1), and some " engineering drawings", which on closer inspection revealed themselves to be glossy adver-tising .)rochures (I W Cochran, TR 2731; I Exh's 6 and 7).
But even the advertising brochures show a containment /
confinement concept for Super Phenix which is the functional
 
                                                                                - F                                                                                          */
equivalent of the CRBRP concept.-                                                      Compare I Exh 7 at 1 and I Exh 5 at 1, with A Exh 1 at 50, TR 2039.
: 36. In spite of the misconceptions which were evident in Dr.
Cochran's characterizations of the Super Phenix design and safety approach, and his lack of any dirset experience in LMFBR design or safety analysis,                                                      Intervenors nevertheless advance the German SNR-300 reactor as evidence that HCDA's should be DBA's for CRBRP.                                                    I Exh 3 at 19, TR 2828. The record, however, does not show that the design characteris-tics, criteria, or state of technology for SNR-300 are in any way similar to CRBRP, or for that matter, what those elements are in the case of SNR-300.
: 37. Dr. Cochran also testified that he was unable to determine                                                        ,
whether the " pattern" established by Super Phenix and SNR-300 holds true for other foreign reactors.                                                      I Exh 3 at 19, TR 2328. The record shows, however, that there is no such
      " pattern".                    The Japanese Monju LMFBR and British CDFR did not treat HCDA's as DBA's.                                                    I W Cochran, TR 2701, 2707; A
!  */  As for the core catcher, the Super Phenix has a provision for retention of the debris from seven subassemblies. I Exh 3 at 19, TR 2828.                                                    This device, however, is not designed to accommodate core-wide disruption, and since it does not seem to be intended as an engineered safety feature to accommodate a complete core melt, this would imply that HCDA's are not DBA's in any real sense. A W Anderson, TR 6369; A Exh 29 at 4; I Exh 5 at 79; see I W Cochran, TR 2649.
  **/
I W Cochran, TR 2597-2604.
 
                                                    - F                Exh 31 at 40-41; A Exh 30 at 7-8; A W Strawbridge, TR 1503-
: 04. On the basis of the record as a whole, it is apparent that the foreign reactor experience cited by Intervenors does not support the proposition that HCDA's should be DBA's for purposes of CRBRP site suitability analysis.
The CRBRP Parallel Design
: 38. Intervenors point to the fact that the original CRBRP appli-cation included two designs:              a) a reference design in which HCDA's were not DBA's, and b) a fall-back or parallel design in which HCDA's were DBA's.              I Exh 3 at 20, TR 2829.                              Inter-venors also point to the Staff's belief at the time of the original application that HCDA's should be treated as DBA's for CRBRP unless and until the Staf f was convinced other-wise.      I Exh 3 at 2G, TR 2829.          Intervenors then recount the history of interactions between the NRC Staff and Applicants from 1974 through May of 1976, when (as Intervenors would have it) there was a " dramatic reversal" of position and the Staff determined that HCDA's should not be DBA's.                                          I Exh 3 at 25, TR 2834.
: 39. Several basic facts in the record shed light on the nature of the Staff's position during the course of its review of the CRBRP application.              First, each and every one of Inter-venors' references to NRC Staff positions indicating that an HCDA should be considered a DBA was dated prior to the date on which the application was docketed.                                        I W Cochran, TR
 
                                                            - F l 2650; A W Clare, TR 1836-37; I Exh 3 at 20-24, TR 2829-33.
It is not surprising that the Staff might adopt a "show me" attitude at least until its review had commenced.                                  Second, the Staff's {{letter dated|date=May 6, 1976|text=May 6, 1976 letter}}, which expressed its deter-mination that HCDA's can be excluded as DBA's, explicitly recognizes that the Staff's determination was its first expression of opinion after having had the opportunity to
                                                                              /
review the application.                                S Exh 5 at 1.-    In light of these facts, the Staff's May 6,                                1976 determination cannot be reasonably viewed as a dramatic reversal, but rather as the Staff's first educated judgment after an opportunity for review of the Application.                                When viewed in this light, Intervenors' historical argument lends no support to a finding that HCDA's should be DBA's for purposes of site suitability analysis.
Role of Quantitative Reliability Analysis in Determining Whether HCDA's Should be DBA's
: 40. Intervenors have maintained that a determination as to whether HCDA's should be DBA's for purposes of site suita-bility analysis must rest first upon definition of a 1
  ""*/  It should also be noted that the design was changed.                                  The Applicants amended the application to revise the containment design from the single steel shell concept in the reference design, to the dual containment / confinement concept now provided for CRBRP.                                S Exh 5 at 4; A W Clare, TR 1837; A Exh 1 at 50, TR 2039.
I
 
                                  - F          quantitative, single-valued reliability objective, and then a demonstration that this objective has been met by detailed, plant specific quantitative reliability and fault tree / event tree (probabilistic) analysis.                '
See, e .~ g . , I Exh 3 at 30-59, TR 2839-68; TR 6623-33, 6639-43.        Intervenors have urged adopting as an objective that the probability of exceeding the 10 C.F.R. Part 100 dose guidelines shall be no greater
                ~
than 10    per year (I Exh 3 at 30, TR 2839), while Appli-cants and Staff have eschewed such an objective in arriving at their respective judgments that the HCDA should not be a DBA for purposes of CRBRP site suitability analysis.          AW Clare, TR 1483; S W Morris, TR 2'278; see S Exh 5.        Nor have either relied upon detailed, plant specific reliability or probabilistic analysis for that purpose.        A W Clare, TR 1483; /  S Exh 2 at 25, TR 2470. Rather, both Applicants and Staff have thoroughly considered and analyzed the pertinent initiators and sequences, criteria, state of technology, and general design characteristics- /    ,  and have applied sound engineering judgment in their evaluations of those con-siderations and in arriving at their conclusions.          S Exh 1 l
  -*/  Applicants' 1975 reliability program considered such a goal, but it has been superseded and Applicants have rejected this approach in favor of a deterministic approach. A W Clare, TR 1483.
  --**/ There is no dispute as to which general design characteris-tics are pertinent. See Findings 10-21, supra.
l I
 
                          - F at II-1-II-20; A Exh 1 at 6-46, TR 1995-2035; S Exh 2 at 4-28, TR 2449-73. Quantitative probabilistic analysis techniques are not sufficiently advanced in today's state-of-the-art to serve as a reliable basis for absolute risk determinations. S Exh 2 at 25, TR 2470.      It is undisputed that this decision must, and properly should, rest upon scund judgment. I W Cochran, TR 2671-79.      Reliance upon the deterministic engineering approach adopted by Applicants and Staff is consistent with longstanding NRC regulatory prac-tice, the weight of the evidence in the record, and common sense. S W Morris, TR 2293; A W Clare, TR 1852; A Exh 1 at 11-14, TR 2000-03. This is not to find that there is no role for reliability analysis in the licensing process for CRBRP . This is to find, however,    t. hat the role envisioned by Intervenors is not sound. The proper role of reliability analysis is not to at priori disprove or prove the proposi-tion that an HCDA should be a DBA through comparison with a single-valued numerical objective, but rather to provide an engineering tool to assure that the potential for high reliability associated with the redundancy, diversity, and independence of the relevant design features is, in fact, implemented and realized. S W Morris, TR 5646-47, S Exh 2 at 25, TR 2470; A W Clare, TR 1689-90.      For these reasons, the Board rejects Intervenors' argument and evidence con-cerning the role of reliability analysis.
 
                                                - F      41. The Board also rejects Intervenors' argument that the Staff's Appendix J analysis (S Exh 8, Appendix J) shows that the HCDA should be a DBA.                TR 6623-33. This argument pro-ceeded from the premise rejected in Finding 40, supra, that one must demonstrate through probabilistic analysis that the design will meet a single-valued quantitative reliability I        objective (see TR 6624) and was predicated upon an inaccu-rate characterization of Staff's and Applicants' testimony (i'.~e., that both Staff and Applicants accept this premise).
Compare TR 6624 with S W Morris, TR 2277-79; S Exh 5; A W Clare, TR 1483.        Having rejected the premise and predicate for the argument, the Board need not even reach the balance of Intervenors' argument.                  In addition, however, the Board would reject the argument for several reasons.                  First, Intervenors' argument presumes that the estimates of HCDA probabilities contained in Appendix J are intended for or applicable to a determination as to whether an HCDA should be a DBA.      Neither Staff nor Applicants have held out these estimates as intended for or applicable to this determina-tion. Appendix J represents a highly conservative or upper bound estimate of the likelihood of HCDA's, and it cannot be meaningfully applied to the determination of whether an HCDA should be a DBA.          S Exh 17 at 8, 14, TR 5755, 5761; A Exh 46 at 12, 21, TR 5388, 5397.                  Second, the balance of the argument itself is based upon an inscrutable series of
 
J
                                - F        Intervenor recalculations of Applicants' and Staff's dose calculations, by which Intervenors argue that these dose calculations are non-conservative by a factor of 10,000. TR 6651-52. These recalculations were presented by Dr. Cochran for the first time at closing argument. TR 6626-33, 6639-
: 43. This argument cannot serve as record evidence (TR 6562-63), and even if decipherable, these calculations have not been sponsored by a qualified witness under oath and subjected to testing on the record. Third, to the extent that the calculations are decipherable, each of the arguments which underlies Intervenors' factor of 10,000 is addressed (and rejected) in the following Findings.
Intervenors' Arguments Relating To The Methods Of Analysis And Assumptions For SSST Analysis
: 42. Intervenors argue that the SSST analysis performed by the Applicants and NRC Staff is inadequate for the reasons that:  a) the HCDA should be a DBA, and thus the SSST and resultant dose consequences should have higher values to envelop all HCDA's (I Exh 4 at 13-18, TR 3063-68) b) the dose calculations should consider the effects of the " entire passage of the cloud" (I Exh 9 at 8-10, TR 3126-28); c) the dose calculations should have assumed the use of a plutonium source term with higher relative concentrations of the isotopes plutonium 238 and 241 (I Exh 9 at 10-14, TR 3128-32); d) the dose calculations did not employ appropriate
 
                                  - F                                  1 dosimetric models (I Exh 9 at 16-20, TR 3134-38); and e) the dose calculations did not include consideration of the
* beyond design basis vent / purge system (TR 6648-51). /
Intervenors' first argument and those advanced in support thereof are rejected on the basis of Findings 10-21 supra.
Each of Intervenors four remaining arguments are considered in the following Findings.
The Entire Passage of'the Cloud
: 43. 10 C.F.R. { 100.ll(a)(2) contemplate      that the dose at the low population zone boundary be calculated for the entire
  --*/
Intervenors' closing argument urged that the doses were underestimated by a factor of 1.5 because the NRC Staff should have used a dose commitment period of eighty years for the maximall: exposed individual, rather than fifty years, which is more appropriate for workers exposed atDr. age twenty. TR 6650-51 (citing I W Morgan, TR 3170-74).
Morgan presumed that the maximally exposed individual would live to be eighty (TR 3170), and by implication, a period of eighty years would be necessary to protect the maximally exposed infant (see TR 3174). Dr. Morgan, however, did not know the average age of individuals living in the vicinity of the site, and thus had no substantial basis for    urging In addition, use of this worst case.      I W Morgan, TR 3174.
Intervenors' argument is of no moment since it mistakes the basic purpose of site suitability analysis. The methods of analysis used by both the NRC Staff and Applicants are based on NRC's standard guidance in NUREG-0172 and TID-14844.          I Exh 9 at 16-17, TR 3134-35; A W Strawbridge, TR 5158; S W Branagan, TR 2344; see 10 C.F.R. { 100.11, n. 1; S Exh 3, Attachment A. The analyses are conducted as an aid in evaluating reactor sites, and the site suitability dose guidelines are reference values against which the analyses can be judged for that purpose, 10 C.F.R. { 100.ll(a) and
: n. 2. The analyses are not conducted to assure radiation protection for the maximaTry exposed individual, nor are the dose guidelines to be considered radiation protection limits 1
for individuals. Id.; S Exh 3, Attachment A at 6, TR 2551.
 
                              - F >                -
period of passage of the radioactive cloud.        While the NRC staff's calculations indicated that the dose contribution after a 30 day period would be negligible, Intervenors nevertheless argue that an NRC Staff sensitivity calcula-tion, which assumed an instantaneous release of the SSST inventory from the containment at the end of the 30 day period, showed doses significantly larger than those cal-culated for the first 30 days. Consequently, Intervenors argue,  the Staff's truncation of their SSST calculations at the end of 30 days results in an underestimate of the SSST doses. I Exh 9 at 9-10, TR 3127-28. Applicants' indepen-dent analysis confirmed the Staff's opinion that the dose contribution after 30 days would, in reality, be negligi-
  ~
ble. Ninety percent of the total 30 day dose would be incurred in the first day, and ninety-eight percent would be t
incurred in the first week. A W Strawbridge, TR 1830-32.
In addition, during the 30 day period of release, the source j
term available in containment for release would be continu-l ously reduced by plateout and fallout of the radioactive materials within the containment. A W Strawbridge, TR 1832. Further, it is unlikely that any accident sequence could maintain the containment pressure at the design pressure for the whole 30 day period, let alone the period beyond 30 days, and thus sustain a design basis leak rate.
I
_-_.-_.v_. -___.-.w.
 
e
                                  - F        A W Strawbridge, TR 1831.      The Staff's sensitivity calcula-tion, however, considered a " puff release" or instantaneous release of all remaining radioactive materials in contain-ment, and did not take into account the effects of plateout S W Bell, TR 2359.
and fallout after the first 24 hours.
When the Staff recalculated the dose effects after 30 days, assuming a reasonable level of fallout and depletion and a design basis leak rate consistent with the Staff's normal SSST analysis (rather than an instantaneous release of all materials), the doses calculated were reduced to an insigni-ficant fraction of the dose calculated for the 30 day peric.i. S W Bell, TR 2400, 2403-04. The " puff release" doses, which are nevertheless within the 10 C.F.R.
                                              */
        } 100.11(a) dose guideline values,- are extremely conservative upper bound values. S Exh 3 at 16, TR 2499;
                                **/ Thus, the NRC Staff's SSST S W Bell, TR 2403-04.---
analysis, which treats the do,se contribution beyond 30 days as negligible, adequately considers the entire passage of the cloud.
See I Exh 9 at 10, TR 3128.
  */
  **/ While
  -~~
Staff Witness Bell's testimony was apparently contra-dictory as to the conclusion to be drawn in this. regard, there is no doubt that the assumptions made in the Staff's puff release sensitivity calculation reflect an undue measure of conservatism, and are not an appropriate    measure of the validity of the Staff's SSST analysis. A W Strawbridge, TR i
1930-32; S W Bell, TR 2353, 2355, 2400-04; S W Hulman, TR 2356-57; S Exh 3 at 16, TR 2499.
l
 
                                  - F    Plutonium Isotopics
: 44. Intervenors argued that the 1% Pu-238 and 5% Pu-241 concen-trations assumed by the NRC Staff for SSST analysis resulted in an underestimation of dose consequences since the rela-tive concentrations of these isotopes in the CRBRP fuel have been underestimated and Pu-238 and Pu-241 are more hazardous relative to the most predominant plutonium isotope (Pu-239).
I Exh 9 at 10-15, TR 3128-3133. Intervenors argue that the concentrations of Pu-238 and Pu-241 are underestimated, since the future use of high burnup LWR fuel or repeated recycle of fuel in CRBRP will result in an increase or buildup of the isotopes Pu-238 and Pu-241 relative to the values assumed in the Staff's analysis. I Exh 9 at 12-13, TR 3130-31. The concentrations of Pu-238 and Pu-241 assumed by the NRC Staff provide a significant margin of conserva-tism in the analysis since the application contemplates values of 0.15% Pu-238 and 2.1% Pu-241.-    A /Exh 14 at 15A-3, 15A-11. In the event that the Applicants should consider the use of fuel with higher concentrations of Pu-238 and Pu-241 in the future, such that the limits of the SSST analysis might be exceeded, the matter would be reviewed by the NRC Staff and a license amendment i
    */  Intervenors' calculated the Pu-241 Value as 0.35% I Exh 9 at 12, TR 3130.
1
 
                                - F                                                                        l
                                                    */
undertaken, as necessary and appropriate.-      AW Strawbridge, TR 1833; S W Hulman, TR 2347. In addition, the Intervenors argument misconceives the nature of the physics characteristics of CRBRP. Unlike an LWR, where the thermal neutron spectrum and repeated recycle would result in buildup of the isotopes Pu-238 and Pu-241, the fast neutron spectrum in CRBRP would cause these isotopes to be burned or reduced in relative concentration with repeated recycling.
A Exh 36, Vol. 3,  14.4A Appendix to Chapter 5.7; I W Cochran, TR 4539. Thus, the Staff's assumed values for plutonium isotopics in the SSST analysis are appropriately conservative, are susceptible to licensing review and amend-ment if and as necessitated by any future changes in fuel characteristics, and are not affected in any significant way by the future prospect of recycling fuel in CRBRP.
Dosimetric Models
: 45. Intervenors argue that the Staff's SSST analyses did not use the most advanced state-of-the art dosimetric models for cal-
!        culating SSST doses and to that extent, the analyses under-l l
l
    -*/  This seems unlikely inasmuch as ample fuel with isotopics
!        within the limits of the analysis is available now. See Finding 171 below.                                  .
 
                                  - F                                                                                      *
                                                                                      /
estimated doses by a factor ranging between 2 and 3.                              The models used are those which are normally applied for this purpose in the NRC regulatory process. I Exh 9 at 16,                          TR 3134, A W Hibbitts, TR 5218. These models have their origins in ICRP-2, and ICRP-2 is, in turn, the basis for the Commission's existing radiation protection standards in.10 C.F.R. Part 20. I Exh 9 at 17, TR 3135; A Exh 25 at 10-11, TR 2084-85. Additional calculations performed by both Applicants and Staff using the more recent models urged by Intervenors showed doses well within the site dose guideline values. S W Bell, TR 2344, A W Strawbridge, TR 1838.
Thus, the models employed are appropriate, the effect of applying more recent models has been adequately
(
  --*/
Intervenors also argued that all transuranic isotopes other than plutonium which are important to dose were not consid-ered in the NRC Staff's analysis. I Exh 9 at 15-16, TR 3133-34. The Staff, in fact, performed analyses of the effect of including other isotopes and concluded that it was small (less than 5%). S W Bell, TR 2340-41.
  ---**/
The significance of the more recent models is relatively slight, a factor of two or so (App Exh 25 at 11, A W Thompson, TR 1904; A W Strawbridge, TR 5156-57, 5282-83; A W Hibbitts, j        TR 5218-19), when considered in light of the fact that the
!        NRC Staff's and Applicants' calculated values for both lung and bone dose are a factor of ten or Smore                        below the site Exh 1 at III-ll; S Exh suitability dose guideline values.
3 at 15, TR 2498; A Exh 1 at 41, TR 2040; A W Strawbridge, TR 1838; S W Bell, TR 2344.
l I
l
 
i
                                    - F
* addressed, /    and that effect in no way invalidates the Staff's or Applicants' SSST analysis and conclusions.
Vent / Purge System
: 46. At closing argument, Intervenors raised an additional point
        -- that the Staff's SSST analysis underestimated the doses by f ailing to consider releases through the beyond design basis vent / purge system. TR 6648-50, 6651. The basic rationale for the argument expressed by Dr. Cochran was that if one considers the system that pumps radioactivity back into the reactor containment building (the annulus exhaust / filtration system), one should also consider the system that pumps radioactivity out of the reactor contain-ment building (the vent / purge system). TR 6649. This argu-ment misconceives the nature and physical function of the proposed CRBRP containment annulus exhaust / filtration system, the SSST analysis, and the vent / purge system. The SSST analysis was properly predicated on the assumption that an HCDA should not be a DBA.      See Findings 10-21, supra.
The SSST analysis assumed that the containment annulus exhaust / filtration system would be available as an
    */    It should also be noted that the site suitability dose
  ~~
guideline values recommended by the NRC Staff include a value of 300 rem for bone surface. S Exh 1 at III-9. In 1977 the Staff had derived a value of 150 rem for average bone using ICRP-2 methodology.      A Exh 25 at 4-5, TR 2078-
: 79. Thus, the dose guideline values are higher by a factor of two for bone surface relative to average bone.
 
                                        - F    engineered safety feature for mitigation of design basis events. S Exh 1 at II-15-II-16, III-10; A Exh 1 at 50-51, TR 2039-40, A Exh 5; A Exh 14. Upon initiation of a release of radioactivity, the containment automatically isolates.
The annulus exhaust / filtration system takes any leakage f rom the reactor containment, passes that leakage through a filtration system, and recirculates that leakage into the containment annulus between the reactor containment building and the confinement building, and not back into the reactor containment building as alleged by Dr. Cochran. S Exh 1 at II-15; A Exh 5.            The SSST analysis considers and accounts for releases through the annulus exhaust / filtration system.          S Exh 1 at III-11; A Exh 14. In the event that conditions progressed beyond design basis accident conditions and containment integrity were threatened during a time period of about a day after initiation of the event, the operator could manually open a normally closed containment vent (the vent / purge system to which Intervenors refer), which would discharge to the environment through a containment cleanup system.          S Exh 1 at II-6; A Exh 1 at 55, 68-72, TR 2044, 2057-61; A Exh 17.          Under design basis conditions, and for purposes of SSST analysis, only the annulus exhaust / filtration system is relevant.        The so-called vent / purge system is closed and has no meaningful physical role under those conditions.        The vent / purge system
 
                            - F  would be manually activated and would release through the cleanup system to the environment only after conditions had proceeded to core disruption, and then only if the contain-ment integrity were threatened about one day after initia-tion of the event.      S Exh 1 at II-15-II-16; A Exh 1 at 68-72, TR 2057-61; A Exh 46 at 29-32, TR 5405-08.      It is therefore neither meaningful nor necessary to consider the ef fects of the vent / purge system in the context of the CRBRP SSST analysis.
l l
I
 
                                            - F  Contention 2e) (Site Suitability Dose Guidelines)
: 47. Intervenors' Contention 2e) alleges that "[als set forth in contention 8[11](d), neither Applicants nor Staff have ade-quately calculated the guideline values for radiation doses f rom postulated CRBRP releases."                Contention lid) particu-larizes contention 2e) by alleging the following:
: 11. The health and safety consequences to the public and plant employees which may occur if the CRBR merely complies with current NRC standards for radiation pro-tection of the public health and safety have not been adequately analyzed by Applicants or Staff.
(d)  Guideline values for permissible organ doses used by Applicants and Staff have not been shown to have a valid basis.
(1)            The approach utilized by Applicants and Staff in establishing 10 CFR Part 100.11 organ dose equivalent limits corresponding to a whole body dose of 25 rems is inappropriate because it fails to consider important organs, e.g.,
the liver, and because it fails to consider new knowledge, e.g., recommendations of the ICRP in Reports 26 and 30.
(2)            Neither Applicants nor Staff have given ade-quate consideration to the plutonium " hot particle" hypothesis advanced by Arthur R.
Tamplin and Thomas B. Cochran, or to the Karl Z. Morgan hypothesis described in " Suggested Reduction of Permissible Exposure to Plutonium and Other Transuranium Elements,"
Journal of American Industrial Hygiene (August 1975).
: 48. The existing NRC regulations specify site suitability dose guidelines of 300 rem for thyroid exposure and 25 rea for whole body exposure.                10 C.F.R. $ 100.ll(a), n.2. The NRC
 
                                                                                                          - F                                                                            Staff developed supplemental guideline values for organs of importance to plutonium exposure for the purposes of CRBRP site suitability analysis.          S Exh 3 at 27-28, TR 2510-11; A Exh 25 at 3-7, TR 2077-81; S Exh 1 at III-9-III-10.          Taking the existing dose guideline values in 10 C.F.R. Part 100 as given, the Staff applied mortality risk weighting factors from ICRP Publication No. 26 (which provide a set of values for the relative radio-sensitivity of the various human organs, based upon a value of unity for the whole body), and scaled from the existing 10 C.F.R. Part 100 whole body and thyroid values to derive two sets of supplemental guideline values for the other organs.          Id,
: 49.                                            The Staff then selected the lowest (most conservative) set of values, which were the values derived from scaling from
                                                                                                                                        */
the 300 rem thyroid value of 10 C.F.R. $ 100. ll( a ) .--      In selecting the ICRP-26 mortality risk weighting factors for scaling, the Scaff reviewed the available state-of-the-art scientific information in the BEIR-I and BEIR-III Reports, and information available from the NCRP and UNSCEAR.          S Exh 3 at 28-2), TR 2511-12.          The mortality risk weighting
                                  --*/                                        Scaling from the 25 rem whole body value in 10 C.F.R. Part 100 could yield a higher set of dose guideline values, including a value of 833 ren for thyroid exposures A Exh 25 at 6-7, TR 2080-81; A W Healy, TR 1893-95; A W McClellan, TR 1901.
 
                                - F      factors of ICRP-26 were the most conservative' values iden-tified by the Staff in its review. Id. In order to account for uncertainty, the Staff' reduced all dose guideline values by a factor of two for the purposes of the Construction Permit review. S Exh 1 at III-9-III-10; S Exh 3 at 30, TR 3513. The approach taken by the Staff in its derivation preserves the validity of the existing 10 C.F.R. I 100.ll(a) dose guidelines, and is entirely consistent with the stated purpose of those dose guidelines:
            . .  . Neither its [the whole body dose of 25 rem] use nor that of the 300 rem value for thyroid exposure as set forth in these site criteria guidelines are intended to imply that these numbers constitute acceptable limits for emergency doses to the public under accident conditions. Rather, this 25                .
rem whole body value and the 300 rem thyroid value have been set forth in these guides as reference values, which can be used in the evaluation of reactor sites with respect to potential reactor accidents of extremely low probability of occurrence, and low risk of public exposure to radiation.
S Exh 3 at 29, T2 2512; A Exh 25 at 3-4, TR 2077-78; see 10 C.F.R. $ 100.ll(a), n.2. The validity of the values derived l
by the NRC Staff for purposes of CRBRP site suitability analysis is well supported by the available body of scien-tific evidence, the existing 10 C.F.R. Part 100 dose guide-line values, and the stated purpose of those dose guideline values.
: 50. Intervenors have argued that more conservative dose guide-line values should be adopted for the following reasons:      a)
 
                                  - F        one should apply the non-stochastic limit of 50 rem per year set forth in ICRP-26, as well as the mortality risk weighting factors set forth in ICRP-26 (I Exh 4 at 28-29, TR 3078-79); b) one should derive the doses by applying weighting factors based upon the EPA environmental radiation protection requirements for normal operation of activities in the uranium fuel cycle (I Exh 4 at 29-30, TR 3079-80);
and c) the doce guideline values should be reduced by a factor of greater than two to account for uncertainties in the dose and health ef fects models (I Exh 4 at 32-33, TR
                  */    Each of Intervenors' arguments is addressed in 3082-83).-
the following Findings.
The Non-Stochastic Limit
: 51. Intervenors argue that the non-stochastic limit of 50 rem per year given in ICRP-26 ehould be applied to derivation of the dose guideline values, in addition to the mortality risk weighting fec tors of ICRP-26.      I Exh 4 at 28-29, TR 3078-79.
Non-stochastic effects are those health effects which show a threshold effect with exposure to ionizing radiation (e.g.,
    */  Intervenors also argued that lower values should be adopted because, in the early 1960's, the ACRS had recommended values of 25 rem for bone and lung.      I Exh 4 at 29, TR
+
3079. Although the record clearly indicates that the ACRS did not recommend this (Compare I W Cochran, TR 2985-2990 with A Exh 33), the Board has ruled that both Applicants' and Intervenors' evidence on this point should be stricken from the record. TR 7102.
e
 
                                                                                                            - F    cataracts), while stochastic effects are those which occur in an apparently random manner and do not show a threshold with exposure to radiation (e.g., cancer).                                                                                    A Exh 25 at 7-8, TR 2081-82.
: 52. The non-etochastic limit corresponds to an annual occupa-tional dose, and its use in deriving dose guidelines would produce values which are higher than those set forth in 10 C.F.R. Part 100 and those derived by the NRC Staff for CRBAP eite suitability analysis (50 rem per year over a 30-year operating lifetime of Clinch River would yield thyroid and lung values of 1500 rem).                                                                                  A Exh 25 at 8, TR 2082; compare S Exh 1 at III-9-III-10.
: 53. Even if the 50 rem per year non-stochastic limit were arti-ficially limited to a one-time exposure, application of the 50 rem per year would require reducing all dose guideline values (except whole body), including the 10 C.F.R. Part 100 300 rem thyroid value, to 50 rem.                                                                                    Id. This, of course, would challenge the validity of the existing regulation in Part 100.                                                  Iji .                                Further, the 50 rem per year non-stochastic limit is designed to limit the incidence of health effects resulting from occupational exposures, snd this purpose is not consistent with the stated purpose of the 10 C.F.R.
      $ 100.ll(a) dose guidelines.                                                                                  (See Finding 49, aupra).
Id. Accordingly. Intervenors' argument for the application
 
                                  - F          of the ICRP-26 non-stochastic limit is neither reasonable nor supported by the evidence in the record.
The EPA Uranium Fuel Cycle Radiation ~ Protection Requirements
: 54. Intervenors argue that the EPA environmental radiation pro-tection requirements for normal operation of activities in the uranium fuel cycle provide a reasonable alternative analogy for derivation of dose guideline values for organs of importance to plutonium exposure.      I Exh 4 at 29-30, TR 3079-80. The EPA requirements contemplate a value of 25 mrem for the whole body, and 25 mrem for any other organ.
Id. Using this analogy, Intervenors argue for dose guide-line values of 25 rem for the whole body and 25 rem for every other organ. Id.
l i    55. Application of the EPA requirements to derivation of the dose guidelines would yield a value of 25 rem for thyroid and result in invalidating the existing 10 C.F.R. Part 100 300 rem thyroid dose guideline value.      Use of this analogy would also ignore the fact that the best available scien-tific evidence shows that the ICRP-26 mortality risk weighting factors describe the relative radiosensitivities of the various human organs in an appropriate fashion.
S Exh 3 at 28-29, TR 2511-12. The ICRP-26 weighting factors ascribe a different and lesser radiosensitivity to all other organs of the human body relative to whole body doses.      A Exh 25 at 6-7, TR 2080-81.
:l
                                                            - F                        56. The bases for derivation of the EPA requirements are incon-sistent with their application as an analogy for darivation of 10 C.F.R. Part 100 dose guidelines.              The EPA requirements included consideration of the health risks attributable to environmental radiation due to the operations of the uranium fuel cycle and the general capability and costs available to 1
mitigate these risks ---i.e., cost / benefit principles.                  39 Fed. Reg. 16906 (May 10, 1974).              There is no evidence in the record to show how the cost / benefit balance was struck by EPA in deriving those requirements, or how that balance incorporates the best available scientific evidence.                  The best available scientific evidence, as reflected in ICRP-26, is most certainly a preferred basis for derivation of the dose guideline values.          Findings 48-49, supra.        In addition, the EPA requirements were intended to " encompass abnormal but anticipated releases of radioactive material to the environment associated with effluent control measures, [but]
potential rel. eases associated with the possibility of acci-dents involving the nuclear safety of the facilities are beyond the scope of the proposed rule making, which is limited to environmental radiation due to normal opera-tion."        39 Fed. Reg. 16906 (May 10, 1974).      (Emphasis added).        Thus, the intent of the EPA requirements is l
entirely inconsistent with their application in the deriva-tion of the 10 C.F.R. } 100.ll(a) guideline values.              On the
 
                                - F        basis of the foregoing, the EPA requirements are not an appropriate basis for derivation of dose guideline values for CRBRP site suitability analysis.
Uncertainties
: 57. Intervenors argue that the Staff's reduction of the dose guideline values by a factor of two to account for uncer-tainties at the Construction Permit stage is nonconserva4 tive. See I Exh 4 at 31-35, TR 3081-85. Intervenors point to the fact that in 1977 the Staff recommended a reduction factor of ten to account for uncertainties, which then included a factor of five to take into account the uncer-tainty in the dose and health effects models. Id. Inter-venors then argued that the uncertainty in the dose and health effects models is still sufficiently high that the Staff's factor of two is not conservative. Id. Intervenors cite three major cources of uncertainty in support of their arguments a) the so-called " hot particle" hypothesis, b) the Morgan bone dose hypothesis, and c) the so-called " warm particle" hypothesis. I Exh 4 at 32-34, TR 3082-84.
: 58. The " hot particle" hypothesis has been considered and rejected by an overwhelming consensus of scientific opinion.
S Exh 3 at 29-32, TR 2512-15; A Exh 25 at 9-10, TR 2083-4; A W McClellan, TR 1916-1920. Intervenors have even stipu-l lated that there is not much support in the written l
l
 
l
                                      - F          literature for the hypothesis beyond that expressed by its
!        authors. TR 6847.
: 59. The Morgan bono dose hypothesis (see I Exh 4 at 32, TR 3082) argues that the maximum permissible body burdens for plutonium 239 set forth in ICRP-2 are non-consertative by a factor cf 240. The dose guideline values recommended by the NRC Staff for organs of importance to plutonium exposure, i
however, were not derived based upon ICRP-2 or the 10 C.F.R.
Part 20 regulations (which were derived from ICRP-2). A Exh 25 at 10-12, TR 2084-86.      The dose guideline values were derived using only the existing 10 C.F.R. Part 100 dose guideline values and the ICRP-26 acrtality risk weighting factors. Id. Consequently, the derivation and validity of the dose guideline values are totally independent of ICRP-2 and 10 C.F.R. Part 20.      Id. Thus, whatever the merits of the Morgan hypothesis, /    it does not affect the validity of
    --*/
The merits of the Morgan hypothesis itself are question-able. The ICRP-30 considered the factors embodied in the Morgan hypothesis, and arrived at body burden values which are a factor of two lower than the ICRP-2 values (as opposed to Morgan's factor of 240).      A Exh 25 at 10-11, TR 2084-
: 85. A W Thompson, TR 1913-1914. Even if the ICRP-30 analysis should lead to a factor of 2 reduction in the 10 C.F.R. Part 20 standards in the future, this would not affect the validity of the CRBRP site suitability dose guidelines, since their derivation ras independent of ICRP-2
!          and 10 C.F.R. Part 20. A Exh 25 at 12, TR 2086.
I          Intervenors also point out that Dr. Johnson challenged the scientific basis for the ICRP-2 values, and by necessary implication, the values in 10 C.F.R. Part 20. See I Exh 21 at 9-11, 12, TR 6026-28, 6029. As with the Morgan hypothe-sis, this evidence is of no moment here, since the validity (Continued) l
 
                                - F        the NRC Staff's recommended dose guideline values.      S Exh 3 at 32-3 3, TR 2 515-16.
: 60. Intervenors also argue that the so-called " warm particle" hypothesis suggests that there is an additional source of uncertainty in the dose guideline values.      I Exh 4 at 32-33, TR 3082-83. Although the record is barren of any evidence to suggest that there is a logical nexus between the " warm particle" hypothesis and the validity of the 10 C.F.R. Part 100 dose guideline values, the record also shows that the
      " warm particle" hypothesis is speculative and not supported by the available scientific evidence.      A W McClellan, TR 4043. Accordingly, the record shows that uncertainties have been adequately considered and that the dose guideline values recommended by the NRC Staff are appropriate for the purposes of CRBRP site suitability analysis. /
of the dose guideline values is independent of ICRP-2 and 10 C.F.R. Part 20. A Exh 25 at 12, TR 2086.
  --*/
Intervenors also point to Dr. Cobb's testimony as an addi-tional source of uncertainty.      I Exh 8, TR 3100-18. Dr.
Cobb's testimony itself can best be described as uncertain. On cross-examination, it was established that Dr. Cobb did not know whether his testimony was relevant to the dose guideline values or not. I W Cobb, TR 2897-98. In fact, Dr. Cobb's testimony addressed the adequacy of the proposed EPA guidelines for contamination in soil. IW l      Cobb, TR 2884-85. These guidelines, which provide screening standards for clean-up of existing contaminated sites, are 1
simply not applicable in any way to the issue of the validity of the dose guideline values.      A Exh 25 at 8-9, TR 2082-83.
l
 
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1
                                      - F              Ultimate Site Suitability Findings
: 61. The CRBRP is a loop-type LMFBR design with a maximum power level of 975 megawatts thermal. S Exh 1 at I-2.
: 62. The CRBRP is intended for use as a demonstration plant for LMFBR technology as part of DOE's overall LMFBR program. S Exh 7 at 'l-1; S Exh 8 at 1-1. It will operate as a baseload unit supplying power to the TVA grid for a five-year demon-stration period. S Exh 7 at 1-1; A Exh 58 at 17, TR 6423.
Thereafter, it will either be purchased by TVA, or continued in operation by DOE, or decommissi7ned. S Exh 7 at 1              1-2.
: 63. The nature and inventory of radioactive materials in CRBRP which are pertinent to CRBRP site suitability analysis have been adequately identified, described, and analyzed.
Findings 23-24, 44, supra.
: 64. The design features which have a significant bearing on the i
probability or consequences of accidental releases of radio-activity have been adequately identified, described, and analyzed. Findings 10-21, 23-24, supra. Certain of these 1
l design features are unique, but all are based upon and incorporate well-proven existing technology. Findings 12-21, supra.
: 65. The safety features engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactivity to the environment can l
l
 
l
                                  - F          occur have been adequately identified, described, and analyzed. Findings 10-21, 23-24, supra.
: 66. The population density and use characteristics of the site environs, including the exclusion area, low population zone, and population center distance, have been adequately iden-tified, described, and analyzed.      S Exh 1 at III-1-III-ll.
: 67. The physical characteristics of the site, which may have a bearing on the consequences of an escape of radiological material from the facility, including seismology, meteorol-ogy, geology, and hydrology, have been adequately iden-tified, described and analyzed.      S Exh 1 at IV-1-IV-7.
Given the hydrological characteristics of the CRBRP site, it is highly unlikely that any significant quantity of radio-active effluent could accidentally flow into nearby streams or rivers or might find ready access to underground water tables. S Exh 1 at IV-2-IV-3. The seismology and geology of the site have been adequately identified, described, investigated and analyzed in terms of vibratory ground motion, the absence of capable surface faults, seismically induced floods and water waves, soil stability, cooling water supply, and distant structures, and the seismic and geologic design bases have been properly determined.      S Exh 1 at IV-3-IV-7.
: 68. Based upon:  a) a postulated fission product release from the core to the containment chat results in potential hazards not exceeded by any accident considered credible; b) i
 
                                                            - F          the expected demonstrable leak rate from containment; c) meteorological conditions pertinent to the site; and d) suitably conservative methods of analysis, the dose to an individual at the exclusion boundary for two hours after the release does not exceed 25 rem to the whole body or 300 rem to the thyroid, or the dose guideline values recommended by the NRC staff for other organs in CRBRP site suitability analysis.                            Findings 10-60, supra.
: 69. On the same basis as stated in the previous Finding, the dose to an individual at the low population zone (LPZ) outer boundary during the entire passage of the radioactive cloud resulting from the postulated release does not exceed 25 rem to the whole body, 300 rem to the thyroid, or the dose guideline values recommended by the NRC staff for other organs in CRBRP site suitability analysis.                        Findings 10-60, supra.
: 70. The population center distance for CRBRP is at least one and one-third times the distance from the reactor to the outer boundary of the LPZ.                            S Exh 1 at III-3.
: 71. Taking into account Findings 6-70, supra, and the evidence in the record as a whole, the proposed CRBRP site is a suitable location for a reactor of the general size and type proposed from the standpoint of radiological health and safety considerations.
u..                                                                            .    . _ _ _
 
l
                                  - F                            ENVIRONMENTAL ISSUES
: 72. There are seven sets of contested factual issues in regard to environmental matters within the scope of NEPA.          These are Intervenors' contentions: A) 2 d), f), g) and h), and 3 c) and d) (environmental effects of severe accidents); B) 4 and 6 b) 4) (safeguards impacts); C) 5 b) (risk to nearby energy and national security facilities due to CRBRP accidents); D) 6 b) 1) and b) 3) (fuel cycle impacts); E) 5 a) and 7 c) (alternative sites); F) 7 a) and b) (likelihood of meeting programmatic objectives and alternative designs);
and G) 11 b) and c) (genetic and somatic effects of CRBRP operation). The Findings which follow address these seven sets of issues and conclude with Findings as to the ultimate I        environmental issues for the CRBRP.
Intervenors' Contentions 2d), 2f), 2g), 2h), 3c) and 3d) (Environmental Effects of Accidents)
: 73. In contentions 2d), 2f), 2g), 2h), 3c) and 3d) (environ-mental effects of accidents) Intervenors allege thet:
: 2. The analyses of CDAs and their consequences by oses Applicants of licensing and  Staff are the CRBR,      inadequate performing      for purp/
NEPA  cost benefit analysis, or demonstrating that the radio-l logical source term for CRBRP would result in potential hazards not exceeded by those from any accident considered credible, as required by 10 CFR 100.11(a), fn. 1.
I
 
                                                                                                              - F                                                                                            d)    Neither Applicants nor Staff have demonstrated that the design of the containment is ade-quate to reduce calculated offsite doses to an acceptable level.
f)    Applicants have not established that the computer models (including computer codes) referenced in Applicants' CDA safety analysis reports, including the PSAR, and referenced in the staff CDA safety analyses are valid.
The models and computer codes used in the PSAR and the Staff safety analyses of CDAs and their consequences have not been ade-quately documented, verified or validated by comparison with applicable experimental data. Applicants' and Staff's safety analyses do not establish that the models accurately represent the physical phenomena and principles which control the response of CRBR to CDAs.
g)    Neither Applicants nor Staff have established that the input data and assumptions for the computer models and codes are adequately documented or verified.
h)    Since neither Applicants nor Staff have established that the models, computer codes, input data and assumptions are adequately documented, verified and validated, they have also been unable to establish the energetics of a CDA and thus have also not established the adequacy of the containment of the source term for post accident radiological analysis.
: 3. Neither Applicants nor Staff have given sufficient attention to CRBR accidents other than the DBAs for the following reasons:
c)    Accidents associated with core meltthrough following loss of core geometry and sodium-concrete interactions have not been adequately analyzed.
d)      Neither Applicants nor Staff have adequately identified and analyzed the ways in which human error can initiate, exacerbate, or interfere with the mitigation of CRBR accidents.
 
                                    - F    74. In the context of matters within the_ scope,of NEPA, the broad issue raised by these Intervenor contentions concerns whether or not the NRC Staff has adequately assessed the environmental impacts of accidents.    (S Exh 8, App. J). The specific point of contest in the record relates to whether the Staff's analysis in Appendix J of the FES (S Exh 8, App.
J), which concluded that the risks of severe, beyond -
design-basis accidents in CRBRP are acceptably low and comparable to those of LWR's (S Exh 8 at J-25), constitutes an adequate analysis of the risks of severe accidents in CRB RP .
: 75. There is no dispute in the record concerning the adequacy of the Staff's analysis of the environmental effects of design basis accidents for CRBRP. An appropriate spectrum of design basis accidents was considered and the realistic dose consequences of such accidents were determined to be: a) comparable to those from postulated LWR accidents, and b)
Well within the dose guideline values for CRBRP site suita-bility analysis.- / S Exh 8 at J-1-J-2; S Exh 7, Table 7.2.
: 76. In regard to accidents beyond the design basis (so-called
            " Class 9" accidents), Appendix J presented an analysis which adequately considered and gave approximately equal attention concerning contention 2 e).
      */    See Findings 47-60, supra,
 
                                            - F              to both the probability of occurrence and the consequences of radioactive releases resulting from inadequate cooling of the reactor fuel and melting of the reactor core. S Exh 8,
" ' " ^
App. J. This analysis also adequately considered:
a) an appropriate range of accident sequences that can lead to releases resulting from core melt (S Exh 8 at J-3-J-12); b) the associated probabilities (Id.); c) the pertinent design characteristics of CRBRP (S Exh 8 at J-3-J-7); d) the risks due to atmospheric and liquid pathway exposures (S Exh 8 at J-12-J-21); e) the environmental consequences of the releases expressed in probabilistic terms (Id.); f) the effects of emergency response measures (S Exh 8 at J-12-J-16); g) the health risks associated with exposures to the surrounding population as a result of the releases (S Exh 8 at J-16-J-21); h) the contribution of external events to the risk (S Exh 8 at J-22); and i) the uncertainties attending the analyses (S Exh 8 at J-22-J-24). The analyses showed that, as with normal operation and design basis accidents, the risks of Class 9 accidents are comparable to those in LWR's. Compare S Exh 8 at J-1-J-2 with S Exh 8 at J-3-J-25.
: 77.      The Staff's analysis estimated the frequencies of severe (Class 9) accidents at the CRBRP involving potential core disruption and containment failure in terms of three phases of such accidents:        a) initiation of core disruption, b) the
 
i I
                              - F      energy associated with core disruption and the resulting response of the primary coolant boundary (with associated releases to containment), and c) the containment response, including the potential for containment failure. S Exh 8 at J-3.
: 78. Based upon LWR experience and the general characteristics of the CRBRP design, the Staff conservatively estimated that
                                                    -4 the frequency of HCDA initiation would be 10 per reactor year or less. S Exh 8 at J-3-J-5; S Exh 17 at 7-14, TR 5754-61; A Exh 46 at 12, 21, TR 5388, 5397. The Staff then considered the potential energy releases associated with core disruption and the resulting response of the primary coolant system, and identified four categories of releases / responses characterized by a spectrum of increasing short-term releases of radioisotopes through the reactor head into containment. S Exh 8 at J-5-J-6. Even though most core disruptive accidents are expected to be nonener-getic and to culminate with an intact primary system with no significant short-term release of radioactive materials through the reactor head to the containment atmosphere, and even though the CRBRP incorporates features to accommodate the energy release from core disruption and limit radio-l active releases from the primary system, the Staff conservatively used head release source terms that imply that all HCDA's would be energetic and would result in the
 
                                    - F        short-term release of substantial quantities of radioactive materials through the head to the containment atmosphere.
S Exh 8 at J-5-J-6, J-8; A Exh 46 at 23-24, TR 5399-5400.
: 79. The Staff estimated the conditional frequency of containment isolation failure, based upon LWR experience and the general
                                              -2 characteristics of CRBRP, a        10    per demand. S Exh 8 at J-6-J-7. The Staff also conservatively estimated the condi-tional frequency of containment failure by overpressure, assuming the failure of all systems provided in the CRBRP to avoid containment failure under core melt conditions, at
          -2                  S Exh 8 at J-6-J-7; A Exh 46 at 29-32, TR 10    per demand.
5405-08. The Staff then defined four classes of HCDA's for analysis, which were based upon appropriate combinations of primary system failure categories and containment failure modes; namely, Class 1 (primary failure category IV and no containment failure), Class 2 (primary failure category IV and containment failure by overpressure), Class 3 (primary failure category III and containment isolation failure),        and Class 4 (primary failure category IV and containment isola-
                                                                        */
tion failure). S Exh 8 at J-8, Table J.2 and Footnote        1.-
    */  Failure Categories III and IV are defined as:
III. Primary system seals experience partial failure A due to excessive mechanical and thermal loads.
limited release of core Pu and other actinides, solid fission products, noble gases, and volatile material into the upper containment would occur immediately. Longer term release to containment
 
                              - F 80. Given the conservative estimates of releases of materials to the containment, the Staff properly estimated the release fractions of radioactive material to the environment, taking into account depletion mechanisms and leakage from the reactor containment building, and calculated the resulting radiological consequences for each of the four HCDA classes. S Exh 8 at J-7-J-12. The Staff considered atmospheric pathway risks, and compared those to the average values of environmental risks for comparable accidents at a typical LWR. S Exh 8 at J-12-J-21. After taking into account the uncertainties in each facet of the estimates of consequences (including early fatalities, latent fatalities, and economic cost) and estimated probabilities, ths Staff concluded that CRBRP accident risks would not be signifi-cantly different from those of current LWR's. S Exh 8 at    J-16-J-25. On the basis of the record, the Staff has cae-quately considered the relevant accident sequences, via the reactor cavity vents would be as in Category II.
IV. Primary system sealing fails open by excessive mechanical and/or thermal loads. A large release of noble gases, volatile material, solid fission products, and core Pu and other actinides could occur immediately. Continuous open venting to the upper containment through failed seals is avail-able for subsequently vaporized sodium and radionuclides. Longer term release to containment via the reactor cavity and vents would be as in Category II.
S Exh 8 at J-5-J-6.
 
l
                                  - F        conservatively estimated the accident probabilities, the releases of radioactive material to the environment, the risks due to the atmospheric and liquid pathway exposures, the economic cost of the loss of the facility, and the
                                          */
effects of emergency measures,- and has adequately addressed the uncertainties in its predictions.        S Exh 8 at J-3-J-25.
: 81. Intervenors have specificially argued that the Staff's Appendix J analysis is inadequate because: a) it has underestimated the frequency of core degradation dua to a loss of heat sink (LOHS) event (I Exh 22 at 14-16, TR 6208-10); b) it has underestimated the frequency of pipe rupture, and erroneously concluded that pipe rupture is not a dominant contributor to the risks of severe accidents (I Exh 22 at 16-22, TR 6210-16); and c) it has underestimated the frequency of containment failure (I Exh 22 at 30-31, TR 6224 /
  */
Intervenors have also argued that the Staf f's CRAC model assumption of LD 50/60 of 510 rads      is unrealistic in regard to supportive medical treatment. I Exh 22 at 34-35, TR 6228-29. The Staff's supportive treatment assumptions          were j
A Exh 55; realistic and consistent with those in WASH-1400.
I W Cochran, TR 6150-51; A Exh 56 at 347; I W Cochran, TR
!        6163; A Exh 57; I W Cochran, TR 6165-68; S W Hulman, TR 5635-36; S W Thadani, TR 5451.
Intervenors have also arg* fed that common mode failures were inadequately considered. I Exh 22 at 22-24, TR 6216-18.
(
    **/
Common mode failures were extensively and adequately                '
l                                                Intervenors argued that considered. See Finding 20, supra.
the frequency of simultaneous failure of both reactor shut-
  .(Continued)
                                                  --                "----*C
                                                          -'-T
 
I
                                - F        The Board's Findings as to each of Intervenors' arguments concerning the inadequacy of the Appendix J analysis follow.
Frequency Of Care Degradation'Due To LOHS Events
                                                  -4
: 82. The NRC Staff assigned a frequency of 10      per year for core degradation due to LOHS. The Staff estimated this value by reference to pressurized water reactor (PWR) reliability experience which indicates that auxiliary feedwater (AFW)
S system failures dominate the frequency of LOMS events.
Exh 17 at 9, TR 5756. PWR auxiliary feedwater (AFW) systems, which are similar to the CRBRP AFW system, show
                                            -4        -5 failure frequencies on demand of 10      to 10    per year.
Id. CRBRP has a backup decay heat removal system (DHRS) down systems had been inadequately considered in light of the Commission's proposed ATWS rule which suggested ATWS frequencies in LWR's on the order of one In perfact, thousand the Staff years. I Exh 22 at 27-28, TR 6221-22.
adequately considered simultaneous f ailure of the two f ast-acting redundant, diverse and independent CRBRP reactor-shutdown systems (as opposed to a single similar system in LWR's) and conservatively estimated the probability of simultaneous failure at 10~4 per reactor year. A Exh 46 at 5-13, TR 5381-89; see Finding 15, supra. Finally, Intervenors have argued in the past that LMFBR's present a unique risk in light of their potential for Intervenors' nuclear explosion. I Exh 3 at 9-10, TR 2818-19.
witness subsequently acknowledged that the physical charac-teristics of a core disruptive accident in an LMFBR are not, as he had previously implied in testimony, similar to those of a nuclear weapon. I Exh 22 at 41, TR 6235; I W Cochran, TR 6154-58. In fact, the physical characteristics of a core disruptive accident are substantially different and more benign than those associated with either a nuclear orFrom a a chemical explosion. A Exh 46 at 25-29, TR. 5401-05.
basic physical standpoint, LMFBR accidents do not involve any risk, much less a unique risk, associated with a nuclear explosion. Id.
 
                                        - F              which does not depend on AFW, so that the decay heat removal function in CRBRP should be at least as reliable, if not more reliable, than that of a PWR. A Exh 46 at 13-21, TR 5389-5397. The effect of artificially establishing the AFW system failure frequency at the high side of the LWR range is to make other potential contributors, such as fuel failure propagation or pipe rupture, at most, small fractions of this dominant contributor to the LOHS frequency. S Exh 17 at 9-14, TR 5756-62; S W Morris, TR 5586, 5590-92.
: 83. Intervenors argued that LWR auxiliary feedwater system reliability studies show higher failure frequencies than those estimated by the NRC Staff. I Exh. 22 at 13, TR 6207. Intervenors cited the Csivert Cliffs auxiliary feedwater reliability study as one example of this, but on cross-examination it was established that this was the only example of which Intervenors were aware.      Id; I W Cochran, i
TR 6110. In fact, the record shows that the Calvert Cliffs AFW system failure frequency is only four times higher than the value estimated by the Staff for CRBRP, even though in contrast to CRBRP, the Calvert Cliffs AFW system is not safety grade, is not automatically actuated, and has a substantially lesser degree of redundancy, independence, and diversity. I W Cochran, TR 6110-21; S W Rumble, TR 5638.
                                                      -4 In addition, the Staff's estimated 10      per year failure
 
                                  - F  ,
frequency for CRBRP is two times higher than that estimated in NASH-1400 for the Surrey reactor.      I W Cochran, TR 6118-
: 20. The record thus shows that the Staff's estimated failure frequency for CRBRP resides at the high end of the range one can associate with LWR experience, notwithstanding the fact the CRBRP design characteristics reflect a substantially higher reliability than those associated with LWR' s . A Exh 46 at 13-21, TR 5388-97; A W Clare, TR 5269; S Morris, TR 5450, 5525; S W Rumble, TR 5559.        LWR experience does not support a higher AFW system failure frequency than that estimated by the Staff for CRBRP, and the Staff's estimate is conservative for application to CRBRP.
: 84. Intervenors also argued that the potential for steam generator leaks and consequent sodium-water reactions could, in fact, control the frequency of LOHS events and render the
                              -4 Staf f's estimated 10 per year failure frequency noncon-I Exh 22 at 14-16, TR 6208-10. In support of servative.
this argument, Intervenors first point out that a steam generator leak could result in a sodium-water reaction, which could in turn produce hydrogen, and raise the The potential for an LOHS event. I Exh 22 at 15, TR 6209.
CRBRP design anticipates this event by providing design features to cope with and limit the consequences of steam generator leaks. The design features include a steam l
generator water side isolation system, a reaction product d
 
                                - F      separator tank, a vent for venting any combustible gases from the steam generator out of the steam generator building, an automatic nitrogen fill system, and an over-pressure protection system. A W Clare, TR 5262-67; SW Becker, TR 6467. Upon cross-examination it was established that Intervenors' Witness Cochran, whose written testimony had raised this argument, was not even familiar with those systems in CRBRP which are designed to accommodate steam generator leaks. I W Cochran, TR 6095-6100, 6160. More-over, the record is barren of any evidence which would suggest that there is a credible mechanism whereby a leak in one steam generator could result in failure of the steam generators in the remaining two loops, or the DHRS. AW Clare, TR 5003, 5006, 5011, 5028, 5267; A W Strawbridge, TR 5017-18, 5020, 5026, 5030.      Thus, it is highly unlikely that steam generator failure would represent a dominant or controlling factor in regard to the loss of heat sink failure frequency.
: 85. Intervenors also point to a GAO letter, which dealt with the Applicants' steam generator test program, as evidence of the fact that steam generator failures may indeed control LOHS frequency. See I Exh 22 at 15-16, TR 6209-10. The problems l
addressed by the GAO letter related to availability con-I siderations, and there is no evidence to show that the letter addresses the effects of steam generator leaks on
 
                                - F        loss of heat sink frequency, or any other safety issue.      See I Exh 22, Attachment 2, TR 6250-60; S Exh 21 at 5-11, TR 6526-32; S W Becker, TR 6467; Findings 216-218, infra.
Moreover, the record shows that the only individual with technical expertise participating in preparation of the GAO letter was a technical consultant to GAO who disagreed with the conclusions drawn in the letter.      I Exh 22, Attachment 2 at 9; I W Cochran, TR 6129-37. Thus, the GAO letter in no way affects the validity of the Staff's estimate of LOHS frequency.
Pipe Rupture' Frequency
: 86. Based upon PWR experience and the design characteristics of CRBRP , the NRC Staff concluded that the frequency of primary pipe rupture would be at most a small fraction of its
_4 conservatively estimated 10    per year LOHS frequency.
S Exh 17 at 12-13, TR 5759-60; S W Morris, TR 5990-91.
Intervenors' Witness Cochran argued that the Staff underestimated the frequency of primary system pipe ruptures and that this frequency may be twelve times higher than that for a PWR. I Exh 22 at 22, TR 6216. It was established I
upon cross-examination that Dr. Cochran is not an expert    in regard to pipe rupture probability. I W Cochran, TR 6100.
Dr. Cochran relied on a report by a Dr. Harris in support of his opinion that the frequency may be twelve times higher than a PWR. I Exh 22 at 19-20, 22, TR 6213-14, 6216. This
 
i
                                - F '
report, however, incorporates the explicit conclusion by Dr.
Harris that the probability of pipe rupture for CRBRP is 0.1-1 times that for a PWR for pipe breaks in CRBRP cf any size (including pin hole leaks). I Exh 22, Attachment 3 at Further, 2,  10, TR 6263, 6271; I W Cochran, TR 6135-36.
subsequent work by Dr. Harris shows that the absolute
                                                                    -7 probability of a pipe rupture in a TWR is in a range of 10
            -8            S Exh 20 at 4-6; I W Cochran, TR 6171-to 10    per year.
: 72. By necessary inference from Dr. Harris' two articles, the probability of pipe rupture in CRBRP would range from
                                                                -4
          -7      -9 10    - 10    per year, as compared with the Staff's 10 estimate of LOHS based on AFW failure frequency.      Inasmuch as Intervenors conceded that Dr. Cochran is not an expert in pipe r?2pture probability, and that they would have to rely on the assertions of Dr. Harris (TR 6694), the weight cf the undisputed evidence suggests that the Staff's estimates of l
pipe rupture probability, based as they are on LWR I      experience, are well supported by the evidence in the record, and that the Staff was correct in concluding that I
pipe rupture would contribute at most a small fraction to its overall LOHS frequency estimate.
Containment Failure Frequency
: 87. Intervenors' Witness Cochran argued that the loss of onsite l
and ofisite power could cause a breach of containment l
l
~
 
f
                          - F  through a loss of containment mitigating systems (princi-pally the annulus cooling and vent / purge systems) and that the Staff had not accounted for this failure mode in estimating the frequency of containment failure. 1 Exh 22 at 30-31, TR 6224-25. The Staff's Appendix J analysis did not assume that any of these systems would be available for about a day after initiation of the event, and its conclusions would not be affected by consideration of this failure mode. S W Rumble, TR 5445-46; S Exh 8 at J-7.        In addition, however, the record clearly shows that the Staff conservatively estimated the frequency of containment fail-ure by overpressure from loss of the mitigating systems. A Exh 46 at 21-23, 29-32, TR 5397-99, 5405-08. In spite of this, Intervenors' Witness Cochran argued that an article in Nuclear Safety shows that the frequency of a breach of containment should be higher by a factor of 10 or more based upon actual LWR experience. I Exh 22 at 31, TR 6224.      Upon cross-examination, however, it was established that while the Nuclear Safety article analyzed the frequency of
  " containment failures", those " failures" were those experienced during leak testing for technical specification compliance in LWR's. A Exh 54; I W Cochran, TR 6147-48.
The leakage requirements embodied in LWR technical specifications are defined as substantially lower. leak rate values (by about a factor of 10) than the design basis leak
 
i
                            - F    rate specified for reactor design basis accident analysis (0.1 volume % per day). Id. In contrast, the Staff's Appendix J analysis estimated the frequency of containment breach or total loss of containment function (S Exh 8 at J-7-J-8), and thus the small leak data in the Nuclear Safety article are simply inapplicable to the failure frequencies estimated by the Staff in Appendix J. S Exh 8 at 12    12-74.
 
1
                                      - F    " Contention 5b) (Ri'sk'to Nearby Energy'and' National'' Security Facilities)
Intervenors' contention 5b) alleges the following:
(
08.
: 5. Neither Applicants nor Staff have established that the site selected for the CRBR provides adequate pro-tection for public health and safety, the environment, national security, and national energy supplies; and an alternative site would be preferable for the following reasons:
b)    Since the gaseous diffusion plant, other pro-posed energy fuel cycle facilities, the Y-12 plant and the Oak Ridge National Laboratory are in close proximity to the site, an accident at the CRBR could result in the long term evacuation of those facilities. Longt,erm evacuation of those facil-it.es would result in unacceptable risks to the 4
national security and the national energy supply.
: 89. The  facilities within the vicinity of the CRBRP which might be affected by a severe accident at CRBRP are the Y-12 Plant, the Oak Ridge Gaseous Diffusion Plant (ORGDP) and the Oak A Exh 47 at 2-4, TR 5422-Ridge National Laboratory (ORNL).
24; S Exh 18 at 4-5, TR 5686-87.        The Y-12 Plant is a major facility within DOE's nuclear weapons production complex and produces various components and subassemblies for the weapons program. The ORGDP enriches uranium for commercial power reactors and conducts development work on advanced isotope separation technologies. ORNL is involved in research related to energy. A Exh 47 at 3-4, TR 5423-24; A W Hibbitts, TR 5273.
 
                                    - F                              ,
l
: 90. Because ORNL is a research facility, rather than a production        ;
facility, long-term evacuation would not significantly impact either national security or national energy supply. A Exh 47 at 4, TR 5424; S Exh 18 at 14-15, TR 5696-97; A    ''t Hibbi t t s ,
In addition, TR 5197, 5244, 5272-5273; S W Sof fer, TR 5657.
because Y-12 plays no role in national energy supply, long-term evacuation of Y-12 would not impact national energy supply. A W Hibbitts, TR 5272; A Exh 47 at 3, TR 5423.
Finally, ORGDP plays no role in national security matters and thus long-term evacuation of ORGDP would not impact national security. A Exh 47 at 3, TR 5424; S Exh 18 at 11, TR 5693.
The inquiry under contention 5(b) is therefore reduced to consideration of the risks to national security in respect to Y-12, and the risks to national energy supply in respect to the ORGDP.
: 91. In considering the impacts of accidents on nearby facilities, both Applicants and Staff conducted an assessment using site suitability source term (SSST) dose calculations. A Exh 47 at 5, TR 5425; S Exh 18 at 6, TR 5688. The SSST was chosen because the consequences of the SSST release are more severe than the consequences of any design basis accident involving a release of fuel and fission products from the core to the containment. See Findings 23-24, supra; A Exh 47 at 5, TR 5425. The SSST thus provides a reasonable bound on the potential ef fects of credible CRBRP accidents upon the
 
                                                                                        ~
4 i
i l
1
                                                                                    - F                                    facilities of interest.                          See Findings 10-22, supra; A Exh 47 at 5, TR 5425.
: 92. In calculating the doses at the Y-12 Plant and ORGDP, Applicants made a number of conservative assumptions which had the effect of overestimating potential exposures at those facilities. First, Applicants used sector specific 5 percent
                                                        */                                                      Second, Applicants meteorology.-                  A Exh 47 at 6, TR 5426.
did not take into acccunt any depletion of the plume.                                                              AW Hibbitts, TR 5219, 5275; S W                                  Soffer, TR 5656. Third, in
.                                    calculating the dose, it was assumed that the person receiving the dose was actually at the.particular facility 4                                    location, but outside any building, twenty-four hours per day 3
for the entire period of the release.                                    A W Hibbitts, TR 5277-78.
                              -*/  Five percent meteorology refers to meteorological conditions for which.'X/Q's are exceeded only 5 percent of the time (also 4
referred L2 as 95 percentile meteorology.) It is in fact very conservative meteorology because environmental                            A Exh                47 assess-  at 6, ments typically use 50 percent meteorology.
TR 5426.
                                **/ In addicion, Applicants made four assumptions. First, wet
                              -~~
deposition, which would have reduced the inhalation dose, was not taken into account, A W Hibbetts, TR 5233-5234; S W Thadani, TR 5656. Second, it was assumed that the entire 30-day release using the SSST occurred over a seven day periodA for purposes of establishing ground contamination levels.
Exh 47 at 8, n.1, TR 5428; A W Hibbitts, TR 5208-5210. A seven day ground contamination dose was used because after seven days the dose would go down rapidly due to decay of        In addition, decontamination iodine. A W Hibbitts , TR 5210.
would have a dramatic impact on dose reduction and thus it would be very difficult to even estimate                                  the dose after seven Third, in calculating deposi-days. A W Hibbitts, TR 5210.
(Continued)
 
                                  - F    93. The doses resulting from the SSST release at the ORGDP are less than both the DOE occupational exposure standard and the Protective Action Guideline (PAG) levels recommended by the Environmental Protection Agency (EPA).              A Exh 47 at 7-9, TR 5427-29; S Exh 18 at 6-7, TR 5688-89.          Although one could evacuate non-essential personnel at ORGDP, production opera-tions would not be affected. A Exh 47 at 7, TR 5427.                                In any event, there would be no long-term evacuation at ORGDP.                                    S Exh 18 at 7, TR 5689.
: 94. The doses resulting from the SSST release at the Y-12 Plant are below both the DOE occupational standards and the Protec-tive Action Guidelines of EPA. A Exh 47 at ' 9-11, TR 5429-31; S Exh 18 at 6-7, TR 5688-89. Because the Y-12 Plant is located 9-11 miles f rom CRBRP, neither short-term nor long-term evacuation of Y-12 wnuld be required. / A Exh 47 at 9-10, TR 5429-30; S Exh 18 at 6-7, TR 5688-89.
tion levels, those elements which were determined to have a
!      minimal impact on doses were not explicitly considered.
Finally, Applicants used the homogeneous core model, which assumes a feed fuel containing 19 percent plutonium-240. The heterogeneous core uses FFTF fuel, which has 12 percent plutonium-240. Using the homogeneous core model has the effect of increasing the dose. A W Strawbridge, TR 5163-5164, 5237. A W Hibbitts, TR 5286. Although the Staff used f
slightly different assumptions, the differences in the cal-culation of doses are not significant. S Exh 18 at 6, TR 5688.
    */    In the event that non-essential personnel were evacuated, the evacuation would only be of short duration during the period of the release, with no significant impact on production. A Exh 47 at 9-10, TR 5429-30.
 
                                                      - F            95. Both Applicants and Staff also considered the effect of an HCDA at CRBRP on the Y-12 Plant and ORGDP in order to provide an additional measure of the risks of CRBRP accidents on nearby facilities.      A Exh 46 at 37-38, TR 5413-14; A Exh 47 at 5, TR 5425; S Exh 18 at 7-8, TR 5689-90.                In considering releases resulting from an HCDA, Applicants analyzed-four HCDA cases and chose the case (Case 2) with the highest radiological releases.      A Exh 46 at 37, TR 5413; A Exh 47 at f
5-6, TR 5425-5426; A W Strawbridge, TR 5181.                  Applicants' HCDA Case 2 is similar to HCDA Class 1 presented in Table J-2 of the Final Environmental Statement Supplement, which was used by the Staff in considering the effects of HCDA's on Y-12 and ORGDP.      S W Thadani, TR 5664; A Exh 46 at 39, TR 5415.
: 96. Applicants' HCDA Case 2 is a representative case for assess-ing the risk associated with beyond design basis accidents on I                          Y-12 and ORGDP. A Exh 46 at 39, TR 5415. Applicants' Case 2 involves containment conditions similar to the Staff's HCDA Class 1. A Exh 46 at 39, TR 5415; A W Strawbridge, TR 5783. More severe events, such as the NRC Staff's HCDA Classes 2,    3, and 4 (See S Exh 8 at J-8) assume multiple failures of design basis features designed to prevent HCDA's, combined with either the failure of containment isolation or the failure of features for mitigating the effects of acci-dents beyond the design basis, such as the containment
 
                                                                                  - F              vent / purge system and annulus cooling system.                                              S Exh 8 at J-                  ,
5-J-8; A Exh 46 at 38, TR 5414; A W Strawbridge, TR 5188.
Such combinations of failures are of exceedingly low probability.                        S Exh 8 at J-8; A Exh 46 at 37-38, TR 5413-14; A W Stravbridge, TR 5187-88; S W Soffer, TR 5664.                                                  In addi-tion, as the successive number of failures increases and the i
consequences of those failures increases the probability of occurrence decreases correspondingly.                                              A Exh 46 at 38-39, TR 5414-15.        Most significantly, the product of the probability and the        consequences (i.e., risk) associated with the Staff's j
HCDA classes, does not increase in progressing from HCDA Class 1 up to Class 4.                                              A Exh 46 at 38, TR 5414. Conse-quently, the risk of HCDA's does not change in progressing from HCDA Class 1 to Class 4, and the Staff's HCDA Class 1 and the Applicants' corresponding Case 2 are therefore repre-sentative and reasonable cases for assessing this risk of HCDA's at Y-12 and ORGDP.                                              A Exh 46 at 37-39, TR 5413-15.
: 97. In analyzing the effect of an HCDA at ORGDP and Y-12, the Applicants used methods of analysis and assumptions similar to those for the SSST analysis, except that 50 percent meteorology, which represents the most probable set of- condi-tions, was used.                                      A Exh 47 at 6, TR 5426.      For the HCDA analysis, additional core particulates (plutonium dominating) were released over an approximately six month period.                                              This was based on the assumption that containment venting and
 
                                        - F        purging is continuous, even though the operator would have the capability to isolate containment and to vent and purge when conditions would be advantageous to dose redu'ction. A Exh 47 at 6, TR 5426.
: 98. The HCDA doses at the Y-12 Plant are below the DOE occupa-tional standard and below the Environmental Protection Agency's Protective Action Guidelines.          A Exh 47 at 11-15, TR
                                                */
5431-35; S Exh 18 at 8, TR 5690.-            Thus, there would not be any significant impact on production at Y-12.            A Exh 47 at 12, TR 5432; S Exh 18 at 8, TR 5690.
: 99. Applicants' calculations of the HCDA release at ORGDP also show that while a short term evacuation of non-essential personnel might be desirable, there would be no impact on production.            A Exh 47 at 11-15, TR 5431-35. In addition,
  -*/  Although it might be desirable to evacuate non-essential personnel at Y-12 during the period of the initial release (the Presodium Boild"y Phase), protective measures would likely be implemented and the radiation doses received by essential personnel would be less than the calculated doses. A Exh 47 at 12, TR 5432.
l
  **/ The
  --~
dosesStaff calculation to the              of the thyroid and    release whole body to ORGDP above      resulted the EPA's    in whole body and thyroid PAG's, and thus the Staff assumed evacuation of ORGDP would be necessary.          S Exh 18 at 7-8, TR 5689-90.
The difference in the Staff's calculated doses and those calculated by Applicants results in significant part from the Staff's assumptions regarding meteorology. S Exh 18 at 8, TR
        $690. Although the wind blows in the direction of ORGDP only approximately 10 percent of the year, S Exh 18 at 9, TR 5691, the Staff, using 10 metet meteorology, assumed for purposes i
of its HCDA analysis that the wind would blow toward ORGDP during the entire period of the release even if the          release Applicants, lasted several days.        S W Thadani, TR 5671-73.
(Continued)
 
l
                                                - F ,                                                                                            ,
I there would be no long-term evacuation at ORGDP and no long-term impact upon production. /
100. The undisputed evidence clearly demonstrates that, in con-4 l              sidering severe accidents at CRBRP, including all accidents within the design base and accidents beyond the design base
( i . e .~ , HCDA's), there would be no significent impact on ORGDP or Y-12 and no significant risk to national energy supply or national security.            A Exh 47 at 16, TR-5436; S Exh 18 at 14-15, TR 5696-97.
101. Intervenors did not introduce any evidence regarding either I                Staff's or Applicants' analysis of the effects of accidents on nearby facilities.            Intervenors have argued, however, J
that both Applicants' and Staff's analyses are inadequate because of " uncertainties", including the following: a)
W Thadani, TR 5671-73.            Applicants, however, used 60 meter meteorology to calculate X/Q's for the total time.            AW Hibbitts, TR 5238-39; S W Thadani, TR 5666.
I
          */    Even assuming a long-term evacuation of ORGDP, there would not be any significant impact on national energy supply.              A Exh 47 at 16, TR 5436; S Exh 18 at 12-13, TR 5694-95.            The    -*
three government owned enrichment plants which provide low enriched uranium fuel (including ORGDP) are now operating at approximately 35 percent of capacity. S Exh 18 at 11-12, TR 5693-94. In the timeframe of CRBRP operations, ORGDP will represent only 18 percent of the total enrich-ment capacity. A Exh 47 at 16, TR 5436. Thus, there is considerable margin for increased operation of the plants (S Exh 10 at 12, TR 5694), and the loss of the ORGDP capacity could be made up at the two other existing plants.          S Exh 18 at 13, TR 5695. In addition, DOE is presently constructing a Gas Centrifuge Enrichment Plant, expected to come on line in 1988, which will provide the necessary capacity to meet U.S. energy requirements into
 
l i
_ y_91 _
wet deposition was not considered; b) more severe HCDA sequences were not considered; and c) doses and releases due to ground deposition were limited to seven days.          Each of these arguments is considered in the following
                  */
findings.-
l 102 Intervenors argued that since wet deposition was not considered, the analysis underestimates the doses. In fact, had wet deposition been considered in the analysis, it would have had the effect of lowering inhalation doses. S W Thadani, TR 5656. This reduction in dose results because both Applicants and the Staff, in calculat-ing doses, assumed no depletion of the radioactive inven-tory in the plume. Rainfall would deplete the inventory of the plume and would reduce the dosage at the relevant loca-tion. S W Soffer, TR 5656.
the 1990's. A Exh 47 at 16    TR 5436; S Exh 18 at 13, TR        >
5695.
    */  Intervenors also argued that recycled fuel was not consid-ered and should have been. This argumentIntervenors is. rejected. See i        Findings 44, supra, and 169-172, infr'a.                    also considered the analyses inadequate because evacuation con-siderations were not taken into account. The Applicants' and Staff's analyses of dose consequences did not take credit for evacuation, and to that extent, they conserva-tive1_y estimated doses. In addition, detailed considera-tions of evacuation are matters appropriately addressed at the CP or OL stage of the proceeding when emergency plan-ning is' considered. Board, TR 5222-29. Finally, Inter-i venors claimed that risk aversion weighting factors should have been taken into account. In fact, the record clearly demonstrates that it would be inappropriate to use such factors. A W Strawbridge, TR 5190-91, 5169; A W Clare, TR 5196.
 
                                          - F      103. Intervenors argued that HCDA sequences with more severe consequences should have been considered.            The HCDA cases selected for analysis of the risk to Y'-12 and ORGDP by Applicants and Staff were, in light of both probabilities and consequences, both reasonable and representative of the expected risk to those facilities from beyond-design-basis accidents.      See Findings 95-96, supra.        Since the risks of HCDA's are low and comparable to those of LWR's, it would not seem reasonable to consider an absolute worst case HCDA for purposes of assessing the probable risks from an envi-ronmental standpoint.          S W Sof fer, TR 5669, 5692. The HCDA doses calculated by Applicants at these facilities are relatively minor fractions of applicable occupational stan-dards, which arovide a significant margin for uncer-tainty. A Exh 47 at 8,      11, 13-15, TR 5428, 5431, 5433-
: 35. Moreover, this margin for uncertainty is expanded by s
the  fact that the doses were calculated assuming: a) no protective measures or dose reduction attributable to evacuation, b) no use of protective devices such as respirators or thyroid blocking agents, c) workers all r
located outside the facilities, rather than the actual l
location of workers within facility structures, d) no use l
of shortened shifts, and e) no reliance upon available l
 
                                                                          - F                                        means  s,  2ntrol accident consequences by operating vent /
purge systems when conditions are most favorable.                                                      A Exh 47 at 6, 7, 12, TR 5426, 5427, 5432; A W Hibbetts, TR 5199-5203; S Exh 18 at 10, TR 5692.                                  Finally, the risk to those facilities-from these highly unlikely accidents at CRBEP has been evaluated by DOE and found to be acceptable from a programmatic standpoint.                A W Hibbitts7 TR 5274.                                        Thus, the cases selected for analysis provide a reasonable and representative basis for assessing the risks to Y-12 and ORGDP, with an ample margin for uncertainty, and with due deference to DOE's programmatic responsibilities.
104.                    Intervenors argued that doses were underestimated because ground contamination levels were calculated for a seven day release period for the SSST case.                                                Applicants calculated ground contamination levels assuming that the entire 30 day release occurred over a seven day period.                                                      The major con-stituents of deposited radionuclides are the short-lived iodine-131 and neptunium-239. A Exh 47 at 8-9, TR 5428-29; A W Hibbitts, TR 5208-10.              The calculated dose would decrease rapidly due to decay, and the dose contribution 1                                        after seven days would not be significant.                                                      A W Hibbitts, TR 5199, 5210.        In addition, decontamination would dramati-AW f
cally decrease the dose after the first few days.
 
                                  - F          Hibbitts, TR 5210. Consequently, the record shows that the seven day release assumption is reasonable. /
      -*/  Applicants' analysis of doses for the four hypothetical accident scenarios ( A Exh 46 at 33-34, TR 5409-10) used realistic gas sparging calculations. A W Strawbridge, TR 5165. This analysis considered c temperature for the pool (4500 degrees F.) which is consistent with the temperature found in the phase of the accident where fuel is melting into the concrete after all the sodium has boiled away. A W Strawbridge, TR 5167. In addition, the analysis considered the dilution of plutonium dioxide by the molten concrete. A Exh 46 at 33 n. 3, TR 5409. Although Intervenors claimed in closing argument that the gas
    .      sparging analyses has a factor of 100 uncertainty    (TR 6797), no citations to the record were given.      In fact, there is no evidence in the record which suggests any uncertainty in regard to the analysis of gas sparging.      The analysis is, in fact, conservative. A W Strawbridge, TR 5171.
 
i i
                                    - F      ,
Contentions 4'and 6 b) 4) (Safeguards) 105. Intervenors' Contentions 4 and 6 b) 4) allege the following:
4 Contention 4 Neither Applicants nor Staff adequately analyze the health and safety consequences of acts of sabotage, terrorism or theft directed against the CRBRP and supporting facilities, nor do they adequately analyze the programs to prevent such acts or disadvantages of any measures to be used to prevent such acts.
a)  Small quantities of plutonium can be converted into a nuclear bomb or plutonium dispersion device which if used could cause widespread deeth and destruction.
b)  Plutonium in an easily useable form will be avail-able in substantial quantities at the CRBR and at supporting fuel cycle facilities.
c)  Analyses conducted by the Federal Government of the potential threat from terrorists, saboteurs and thieves demonstrate several credible scenarios which could result in plutonium diversion or releases of radiation (both purposeful and acci-dental) and against which no adequate safeguards have been proposed by Applicants or Staff.
d)    Acts of sabotage or terrorism could be the ini-tiating cause for CDA's or other severe CRBR acci-dents and the probability of such acts occurring has not been analyzed in predicting the proba-bility of a CDA.
Contention 6.b.4:
i The impact of an act of sabotage, terrorism or theft directed against the plutonium in the CRBR fuel cycle, including the plant, is inadequately assessed (as regards environmental impact), nor is the impact of various measures intended to be used to prevent sabotage, theft or diversions.
Board Order of April 14, 1982, Appendix I at 5-6, 9.          .
x
 
                                              - F            106. The thrust of these concentions is that the risks and costs associated with safeguards for CRBRP and its fuel cycle have not been adequately assessed.          The CRBRP will be licensed by the NRC and subject to NRC safeguards requirements.                The fuel cycle facilities for CRBRP will be owned by DOE and subject to DOE safeguards requirements.                S W Dube, TR 3719-          ;
3720; S Exh 7 0 at 12, TR 3744.
107. DOE threat guidance requirements for like materials are as high or higher than the counterpart requirements of the NRC. S W Jones, TR 3620-3621, 3627.            This applies to all categories of threats, inc'.uding external assault coupled with an insider or insiders, theft by an insider or insiders, and external force threats.              S W Jones, TR 3627-3635.        Safeguards designed in accordance with DOE's require-ments will provide a level of protection against theft and sabotage that is at least as high as that provided by safe-guards designed in accordance with NRC's requirements. S Exh 8 at E-3; A Exh 35, Vol. 2, Chapter 5.7 at 5.7-41; S Exh 10 at 12, TR 3744.
108. DOE has ongoing, effective safeguards programs to assess threats, / evaluate the effectiveness of safeguards at
            --*/
Threat assessment is used to provide a clear picture of the potential adversaries, their        capabilities and motivations, A Exh 39 at 7-9, TR 3480-82 Both NRC and their objectives.
and DOE have systematic threat assessment programs              to review S W Jones, TR possible changes in the design basis threats.
3717-3718; A Exh 39 at 7-9, TR 3480-82. Although changes in (Continued)
 
I
                                  - F                              */                                    **/
existing facilities- and improve safeguards technology.
A Exh 39 at 7-16, TR 3480-3489; A W Hammond, TR 3302, 3421; A Exh 35, Vol. 2, Chapter 5.7 at 5.7-41. NRC, DOE and DOD continually exchange information from ongoing safeguards research programs. S W Hockert, TR 3723. The results from these various programs, including new technologies, are and will be made available to CRBRP and its fuel cycle facili-ties. A Exh 39 at 11-16, TR 3484-89; A W Hammond, TR 3455, 3460, 3464.
4 109. The ongoing DOE safeguards programs have established an extensive technological base for design, installation, operation, and maintenance of effective, in-depth, physical the threat level of 1 to 2 persons might not be detected, such a change would not lessen the effectiveness of the safeguards system. A W Penico, TR 3423-24. In order to affect the safeguards system, the change would have to be on the order of 5-10 persons.      Such changes in the threat can be detected by the intelligence organizations. A W Penico, TR 3424-3425. In addition, in the event of a change in perception of the threat, NRC can issue an immediately effective order to upgrade security requirements to meet the threat change. S U Jones, TR 3718. Such an order could be issued overnight as a license condition. S W Dube, TR 3687.
  */    Both NRC and DOE have developed methodologies and approaches to evaluate the effectiveness of safeguards systems. A Exh 39 at 9, TR 3482; A W Hammond, TR 3430-3432. These metho-dologies include the use of fault tree and decision analysis and " black hatting" both of which have been used at CRBRP.
A Exh 39 at 10-11, 32-34 TR 3483-3484, 3506-3508; A W Penico, TR 3460-3461.
    **/  An important element of DOE's safeguards program is the
  ~~-
development and improvement of technology fdr physicalAfter prctection and material control and accountability.
development and testing, these technologies will be avail-able to be incorporated in the safeguards system design.
A Exh 39 at 11-12, TR 3484-85; A W Hammond, TR 3464-3465.
 
I
                                                    - F                                                                              a protection systems in support of the CRBRP and its fuel cycle facilitiers.        A Exh 39 at 12, TR 3485; A W Hammond, TR
                                */
3302-03.-
CRBRP~' Safeguards 1
110. Due to the inherent design characteristics of the CRBRP and its fuel handling system,            theft of plutonium is a highly unlikely event, irrespective of the physical security sys-tems which have been designed for the plant.                A Exh 39 at 23, TR 3497.          The fuel assemblies containing plutonium in oxide form will be delivered in single assembly con-tainers.        The containers and the fuel assemblies weigh approximately three thousand pounds.          A Exh 39 at 23-24, TR 3497-98.        Each individual assembly itself is 14 feet long
              */      Budgetary constraints have not precluded the introduction  AW of safeguards technologies into specific facilities.
Hammond, TR 330A-05.
              **/ Accountability            of fissile and fertile material is inherent in the design of the CRBRP refueling system. After inspec-tion at receipt, the assemblies are not visually identified again until shipment of the irradiated assemblies. The
'                    assemblies are mechanically identified prior to insertion into the core and subsequent to removal from the core as part of the refuelino controls. All movements of fuel within the plant are monitored and/or recorded on the.
refueling system computer for inventory purposes and 2,      to insure proper configuration changes.          A Exh 35, Vol.
Chapter 5 at 5.7-65; S Exh 8 at E-10.
                ***/ As conceded by Intervenors, irradiated fuel is not an attractive or even likely theft or sabotage target (TR 3252) because it is both radiologically and thermally " hot".          A Exh 39 at 68, TR 3542; S Exh 10 at 10, TR 3742; S W Dube, TR 3573.
1
 
l
                                  - F ,
and weighs approximately 450 pounds.                        These 450 pound assemblies remain as assembled units during their entire life at the CRBRP plant.            A Exh 39 at 24, TR 3498.
111. Except for initial inspection and final preparation for shipment, the fuel assemblies are stored in massive tanks at a temperature of over 400 degrees F. in molten sodium and under an inert atmosphere.              A Exh 39 at 25-26, TR 3499-3500:
A Exh 35, Vol. 2, Chapter 3.8 at 3.8-4.                        All fuel handling operations are under continuous closed circuit television coverage and are performed remotely, or with substantial shielding around the assembly. A Exh 39 at 24-26, TR                                  3498-3500. Guards will be present whenever fuel is moved.                                    A Exh 39 at 26, TR 3500. The inherent security at CRBRP makes theft of fuel a highly unlikely event.                          A Exh 39 at 26, TR 3500.
112. The fortress-like nature of the plant, with walls up to six feet thick, and location of individual components in l        separate reinforced concrete cells, provides substantial inherent protection against sabotage.                        A Exh 39 at 27-28, TR 3501-3502. Sabotage of the CRBRP would only be "possible" if all safer srds were stripped from the plant and two well-qualified insiders were given unlimited and uncontrolled access to the plant.      A W Penico, TR 3255, 3258.                          Multiple layers of controls have been incorporated directly into the plant design which will substantially minimize the likeli-t
_ _ - _ _  ,_.__v y    - - - --,
 
                                    - F-100 -
                                              */  A Exh 39 at 28-29, TR hood of radiological sabotage.-
3502-3503. The principal events that could lead to an HCDA and thus potentially result in radiological releases are excessive power generation or reduced heat removal events without scram. A Exh 39 at 28, TR 3502. These events could only result through multiple system failure.                    A Exh 39 at 28, TR 3502. Access to these systems is limited to autho-rized personnel only.            Additionally, detailed knowledge of the design and operation of the plant protection system, control system, and hardware would be required.                  A Exh 39 at 28-29, TR 3502-03. Any deliberate attempt to initiate a transient would require manipulations of complex electronic / electrical circuitry, with small margin for error. Any mistake by an adversary in manipulating the plant protection systems could result in reactor scram.                          A Exh 39 at 31, TR 3505; A W Penico, TR 3262-3263, 3444-3445. Moreover, in order to sabotage the plant, a saboteur would have to have access to at least two vital areas con-currently (i.e. he would have to be in two places at the A W Penico, TR 3283-84.            In addition, the plant same time).
systems are equipped with sensors which will alarm at any attempt to place the plant in any unsafe or abnormal
*/    In discussing sabotage, the relevant area of interest See                    is    A
~
radiological sabotage as defined in NRC regulations.
Exh 39 at 27, TR 3501; A W Penico, TR 3285.
 
1 l
                                    - F-101 -
condition. A Exh 39 at 28, TR 3502.              The plant design also contains a manual control system in the control room with widely separated manual scram buttons, thus permitting manual scram upon indication of an unsafe condition. A                Exh 39 at 30, TR 3504. The    inherent design features of the plant, including the fuel handling system, and the indepen-dent, diverse and redundant safety features make theft and sabotage highly unlikely events even without consideration of the physical security system.      A Exh 39 at 23-32, TR 3497-3506.                                                              l CRBRP Physical Security System 113. The security system for CRBRP incorporates the most advanccl analytical techniques and technology.                The analytical efforts included vulnerability analyses, location analysis and criti7al path analysis.      These efforts, which also included black hatting exercises and fault tree analysis, provide additional assurance that the safeguards design will be effective against sabotage or theft. A Exh 39 at 32-36, TR 3506-10; A W Penico, TR 3466-3467 114. Four security areas with increasingly stringent security will be designated:    (1) Controlled Area; (2) Isolation Zone; (3) Protected Area; and (4) Vital Areas.              A Exh 39 at 39, TR 3513.
 
l
                                      - F-102 -
115. The' Controlled Area, which includes the owner controlled area outside the security barrier, will.be marked by signs or other means to ensure that persons entering the area are aware they are on private property.                Patrol roads will facilitate locating and removing persons from this area when required.      A Exh 39 at 41, TR 3515.
116. The Isolation Zone, which is an area straddling the fence line, is cleared of all obstacles which would impede vision.
It is roughly 30 feet outside and 10 feet inside the fence. A Exh 38 at 41, TR 3515.
117. The Protected Area, which is an area within the Controlled Area, will be completely enclosed by a security barrier through which controlled access is strictly enforced.                    All 4 structures and components necessary for the safe operation of the CRBRP are within the protected area security
                                                                            */
barrier. A Exh 39 at 41, TR 3515; S Exh 8 at E-9.--
      -*/  The physical security systems associated with safeguarding the protected area include proper grading the landscaping to facilitate maximum visual and closed circuit television monitoring; lighting; security barrier fence; multiple, sectionalized intrusion-det2ction systems located on and along the security barrier fence; perimeter patrol            road; and These a closed circuit television monitoring system.
systems not only will deter threats, but also will alert personnel in the Central and Secondary Alarm Stations when an external threat exists. The trained on-site and off-site guard force and local law enforcement agencies can then be contacted from either of the continuously manned Central or Secondary Alarm Stations. A Exh 39 at 42, TR 3516; S Exh 8 at E-9.
o
 
                                - F-103 -
118. Vital' Areas contain vital equipment and receive maximum protection and access control.      All vital areas associated with the CRBRP are located within the fenced and alarmed protected area. A Exh 39 at 41-42, TR 3515-3 516.
119. Access through the protected area barrier is controlled by security guards located at the Access Control Station.
Physical search will be aided by explosives detectors and metal detectors to enable detection and identification of explosives and firearms.      A Exh 39 at 43, TR 3517. Access within the plant structures is controlled by computer based card readers to ensure that no personnel are granted unauthorized access. A Exh 39 at 43, TR 3517; A W Penico, TR 3468. This is administratively supplemented by personnel screening and monitoring, a photo-identification system, escorts when required, and control of personnel traffic flow. A Exh 39 at 43, TR 3517.
120. Access to vital areas is even more stringently controlled.
Approximately 2-3 percent of the plant personnel will have access to all vital areas and no personnel will have uncon-l      trolled access. A W Penico, TR 3279-3280; A Exh 39 at 44-45, TR 3518-19. Access to vital areas is based strictly on necessity. A W Penico, TR 3280. The vital areas are separated by function, and vital equipment and systems are located in inerted cells and spaces not accessible during normal operation. A Exh 39 at 44, TR 3 518. Entry to vital
 
i i
                                                      - F-104 -
areas is controlled by a dual computer based card reader system which continuously monitors the status of all vital area doors. Alarms will nound in the event a door remains open too long, control wiring is cut or a door is forced open. A Exh 39 at 44-45, TR 3518-19; A W Penico, TR 3285-
: 86. /
121. The CRBRP will be equipped with redundant and separate communication systems to provide communications en-site r
                      -  between security stations and guard force personnel, and off-site from the Central Alarm Station and Secondary Alarm Station.      A Exh 39 at 44-45, TR 3519-20. S Exh 8 at E-9.
The Central Alarm Station and the Secondary Alarm Station utilize redundant and independent computers which are complemented by the on-site security force.              A Exh 39 at 46, TR 3520; S Exh 8 at E-9.
122. The CR9RP will institute a vigorous screening process of all plant employees. All employees will undergo physical exami-nation by a licensed physician, security investigations, a National Agency check, and psychiatric examinations will be given when either the examining physician believes it neces-sary or when an employee's performance indicates the need for such an examination.        A Exh 39 at 36-38, TR 3510-3512; A W Penico, TR 3270-3274, 3375-3376.
                    */    The cards used in the card reader system are not susceptible to duplication.      A W Penico, TR 3290.
 
                                - F-105 -
123. Advanced analytical techniques will continue to be applied to the design to test the system against a wide range of threats as they may evolve. A Exh 39 at 48, TR 3522; A W Penico, TR 3267-3268, 3426. The advanced design of the CRBRP security system, including its modular design, pro-vides flexibility for accommodating future design changes in a timely manner, if that should be necessary.      A Exh 39 at 48-49, TR 3522-3523; A W Penico, TR 3426-3427.
Cost of CRBRP Safeguards 124. The capital cost of engineering and installing an effective security system that meets all regulatory requirements is about $3.8 million. A Exh 39 at 49, TR 3523; A Erh 35, Vol.
2, Chapter 5.7 at 5.7-64; A W Penico, TR 3394-3395. CRBRP security operating costs are estimated at under $2.5 million per year during the demonstration period. A Exh 39 at 50,
                                  */
TR 3524; S Exh 8 at E-lO.--    The modular design of the security system will allow improvements to be made with small or no cost impact. Improvements in state of the art technology, particularly in the area of microprocessors, may well reduce capital costs. A Exh 39 at 49-50, TR 3523-3524; A W Penico, TR 3399-3403.
--*/
The cost estimates for safeguarding CRBRP and its fuel cycle facilities were independently analyzed by Battelle Northwest and found to be reasonable. S W Dube, TR 3668-3669.
 
                                - F-lO6 -
125. It is undisputed that the planned CRBRP safeguards system for CRBRP will exceed NRC licensing requirements. AW Penico, TR 3452.      In addition, Applicants are committed to implement an effective safeguards system, irrespective of
                      ~
the NRC regulations.      A Exh 39 at 23, TR 3497; A W Penico, TR 3451-52. There are no factual disputes with regard to the ef fectiveness of the safeguards system planned for CRBRP or the inherent protection provided by the plant design.
The safeguards requirements, inherent plant design charac-teristics, the physical security system and the MC&A system planned for CRBRP make the likelihood and risk of thef t or radiological sabotage extremely low.      A Exh 39 at 22, 79, TR 3496, 3553; S Exh 10 at 12-13, TR 3745; S Exh 8 at E-lO-E-
: 11. The economic costs of safeguarding CRBRP against thef t and sabotage are, and are likely to remain, small (less than 1%) fractions of the total plant cost. A Exh 39 at 48-50, TR 3522-3524; A W Penico, TR 3399-3403.
Fuel Cycle Safeguards 126    The CRBRP fuel cicle includes mixed oxide (MOX) fuel fabri-cation, blanket element fabrication, reprocessing, manage-ment of the wastes generated by the various facilities, and transportation of wastes and products among the various A Exh 39 at 16, TR 3489; A Exh 35, Vol. 2, facilities.
Chapter 5.7 at 5.7-40. All DOE CRBRP fuel cycle facilities will implement safeguards systems, consisting of a physical
 
                                                            - F-107 -
security system and a MC&A system, in accordance with DOE Orders 5632 and 5630.                              A Exh 39 at 50-77, TR 3524-3551; A Exh 35, Vol.                      2, Chapter 5.7 at 5.7-41-42; A W Hammond, TR 3307-3309. /
Fuel' Fabrication 127. CRBRP fuel pins will be fabricated in the Secure Automated                                                                                                    ,
Fabrication (SAF) line now under construction at the Fuels and Materials Examination Facility (FMEF) at Hanford. A Exh 39 at 51, TR 3525.                              The FMEF/SAF will be protected by an integiaded safeguards system composed of physical security and material control and accountability.                                                        A Exh 39 at 51, TR
    --*/
The applicable DOE orders provide a level of protection comparable to the NRC regulations. See Finding 107 supra.
The two subparts of the safeguards system are complementary                                                                                                  ,
and are not intended to be independent, stand-alone systems.
A W Hammon3, TR 3363, 3432. Although not required by NRC regulations, the material control and accounting systems for the fuel fabrication and reprocessing facilities will use the most advanced technology for remotely controlled auto-mated processing and near real time accounting                                                      techniques.
A Exh 35, Vol. 2, Chapter 5.7 at 5.7-42. Reliable, accuzate non-destructive assay (NDA) techniques for determining the content of uranium and plutonium have been installed and successfully tested on existing DOE facilities. A Exh 39 at 15, TR 3488; A W Hammond, TR 3335. The NDA techniques have been coupled with near real time analyses methods to provide                                                                                                '
continuous monitoring of changes in the amount of SNM.in the particular facility. A Exh 39 at 15, TR 3488; A W Hammond, TR 3339-3345; S W Dube, TR 3688-3689, S W Hurt, TR 3690-3691. The near real time accounting systems components have been thoroughly tested and are available for use in DOE's fuel cycle facilities.                                A W Hammond, TR 3335-3339.                          SW Dube , TR 3688-3690; S W Hurt, TR 3690-3691. A Exh 35, Vol.
2, Chapter 5.7 at 5.7-58.                                            A near real 'ime accounting system is not required by NRC regulativas. S W Dube, TR 3688.                                                                                            .
                    . . _ . _ _ . _ _ .            ,, ._    .r_  , , _ - - . _
 
I
                                            - F-108 -
3525.        The system employs physical barriers around the protective area with armed guards and intrusion detectors.
l    A Exh 39 at 54, TR 3528. The protected area is illuminated and under constant closed circuit tels/ision (CCTV) surveil-lance.                A Exh 39 at 55, TR 3529.            Access to areas and struc-tures within the protected area is strictly controlled and limited by the intrusion detection, entry control and internal surveillance systems.                          These systems employ the best available components and techniques, including hand geometry identification, TV displays, electrically locked doors, computer data processing and data analysis.                                                    A Exh 39 at 55, TR 3529.                In addition, the system is modular to allow for installation and evaluation of advanced safeguards equipment and systems.              A Exh 39 at 51, TR 3525; S Exh 8 at E E-7; A Exh 35, Vol. 2, Chapter 5.7 at 5.7-43-44.
128. Complementing the physical security system is the materials control and accountability system which is carried out on the Safeguards Computer Operating System (SACOS).                                              A Exh 39 at 52-53, 55-56, TR 3526-27, 3529-30.                                This computer system l
operates in a near real time mode through direct links to
    ' the process control computer and can detect diversions of special nuclear material within hours. A Exh 39 at 55-56, TR 3529-3530; A W Hammond, TR 3340-3345.                                Materials moving through FMEF/SAF are continually monitored and measured using NDA, chemical analysis and laser scanning.                                            A Exh 39
 
i 1
l l
                                                          - F-lO9 -
at 52-54, TR 3526-28; A Exh 35, Vol.                          2, Chapter 5.7 a- 5.7-46-48. The SACOS system is protected from unauthorized access through securs communication wireways, by limiting individual access and through the use of hand geometry idantification.                  A Exh 39 at 55-56, TR 3529-30; S Exh 8 at E-7-E-8.        The integrated physical security system and MC&A system at FMEF/SAF assure that the risk of theft or sabotage are acceptably low.                A Exh 39 at 51, TR 3525.
* 129. The CRBRP fuel cycle will utilize about 65 percent of SAF's operational schedule, and that portion of the safeguards costs are applicable to CRBRP. Safeguards costs for fabri-cation of CRBRP fuel assemblies are:
Initial costs:                            $1.5M Annual cost:
Repair & Maintenance                      $ .2M Guard Force                                $ .6M Total Annual Cost                        $0.8M A Exh 39 at 56-58, TR 3530-3532; A Exh 35, Vol.                        2, Chapter 5.7 at 5.7-61-5.7-63.
Reprocess.ng i 130. Although it is anticipated that CRBRP fuel will be repro-cessed in the planned Developmental Reprocessing Plant (DRP), two alternative facilities were also considered:                        a) a small facility dedicated exclusively to CRBRP and FFTF
 
                                    - F-110 -
fuels with approximately 15 tons per year capacity; and b) a breeder fuels head-end capability add-on to an existing LWR fuels reprocessing plant. A Exh 39 at 21, TR 3494.
131. The small dedicated facility can be adequately and easily safeguarded with routine application of existing safeguards    '
technology. A Exh 39 at 21a, TR 3495. Effectively safe-guarding higher capacity facilities, such as DRP, will require more extensive safeguards systems with higher costs, and thus the costs of the DRP safeguards bound those for the
                          *                                  **/
alternatives. /  A Exh 39 at 21a, TR 3495.
132. The reprocessing activities planned for CRBRP fuels are essentially comparable to the activities now ongoing in existing DOE program and facilities. A Exh 39 at 71-72, TR 3545-3546. Effective safeguards monitoring techniques and analytic methods for these activities and ongoing technology development programs are in place. A W Hammond, TR 3405-3406;  Substantially equal throughputs of plutonium to those planned for DRP are now achieved and effectively safeguarded in the U.S. military programs. A Exh 39 at 75-76, TR 3549-3550; A W Hammond, TR 3436-3437.
    --*/
Ofcourseanadd-ontoanexihing,tadilitycoulduseexist-ing safeguards, and thus 'would involve a less'er cost than an all new facility such as DRP.
      **/    Applicants are comr.itted to impledenting effective safe-guards at the reprocessing f acility regardless of the facility.used.- A Exh 35, Vol. 2, Chapter 5.7 at 5.7-40; S W
                              ~
Dube, Th 3683-3684.
                                      ~~
w  ..
                          ''                                      t
 
                  =  .-
t l
l l
[
'                              - F-lli -
l                                      .
133. The DRP's physical protection system is designed to protect I      nuclear materials from theft or diversion through the use of access and egress controls and physical barriers, surveil-lance measures and alarm systems, and on-site security personnel and off-site law enforcement assistance. A Exh 39 at 72-73, TR 3546-3547; A Exh 35, Vol. 2, Chapter 5.7 at 5.6 5.6-55 ; S Exh 8 at E-ll - E- 12. The physical protection system design includes: SNM contraband detection components, forcible entry detection components, covert entry detection components, damage control procedures, communications systems, armed response forces, and personnel surveillance. Each of these physical protection systems elements is an integral component of the entry-control and intrusion detection subsystems and the safeguards response and control system. Design basis criteria and guidelines for hese subsystems have been prepared. A Exh 39 at 73, TR 3547; A Exh 35, Vol. 2, Chapter 5.7 at 5.7 5.7-55; S
!      Exh 8 at E-ll - E-12. The DRP will rely extensively on remote operations and maintenance procedures, thus limiting the access to materials, and minimizing the possibility of diversion or sabotage. A Exh 39 at 72, TR 3546.
134. Advanced MC&A techniques which have been effectively tested and demonstrated will be incorporated in the DRP. A Exh 39 at 73-75, TR 3547-49; A W Hammond, TR 3383, 3404, 3446. A near-real-time accounting (NRTA) demonstration at the l
l i
 
                                                                                    \
l
                                    - F-ll2 -
Barnwell Nuclear Fuel Plant (BNFP) shows that NRTA can l
l        sic.ificantly increase the sensitivity and timeliness of diversion detection relative to conventional accounting to A Exh 39 at 73-74, TR 3547-permit detection within hours.
                                        */ See note accompanying Finding 3548; A W Hammond, TR 3339.-
126 supra.
135. The initial cost of DRP safeguards would be about $50 A
million. Operating costs are estimated at $12.5 million.
Exh 39 at 76-77, TR 3550-3351.        CRBRP will use about 8 per-cent of the DRP's capacity, and the pro-rata cost of CRBR safeguards is about 4 million dollars (capital) and about 1.1 million dollars (annual operating). A Exh 39 at 76-77, TR 3550-3551.      If the facility option selected for repro-cessing of CRBRP fuels is a low throughput dedicated facility, effective safeguards can be applied at costs comparable to or less than the pro-rata costs described above. A Exh 39 at 77, TR 3551.
Transportation 136. There is no apparent dispute in the record concerning safe-guards for transportation. The DOE Transportation Safeguard In addition to NRTA, process monitoring which recognizes
    ~~*/  significant patterns of changes in on-line measurements has been successfully demonstrated. A Exh 39 at 74, TR 3548.
l          Also, penetration monitoring and monitoring techniques for operational equipment and maintenance equipment are under developme3t. A Exh 39 at 74-75, TR 3548-49.
 
l l
l
                                            - F-ll3 -
System (TSS) is planned for use in transporting fresh MOX fuel assemblies and spent          f uel .- / The system serves approximately 125 shippers and receivers of SNM and other sensitive materials at approximately 100 locations through-out the United States and provides weapons-level protection to all such shipments.          A Exh 39 at 58-60, TR 3532-3534. At.
the present time, the DOE TSS ships quantities of plutonium equivalent to the quantities which will be generated by CRBRP .        A W Hammond, TR 3417. The system is an effective combination of specially designed transportation equipment, nationwide communications, and armed couriers which assures that the risks associated with safeguards transportation are extrGualy low.        A Exh 39 at 60-64, TR 3534-3538; A Exh 35, i          Vol.        2, Chapter 5.7 at 5.7-48-5.7-49.
l 137. The transportation cost of fresh CRBRP fuel will be but a small incremental increase to the already existing transpor-tation system.        The incremental cost is expected to be less
    --*/
In transit, spent assemblies would be procected in large casks weighing many tons to minimize radiation. Irradiated assemblies would be contained in a removable canister-inserted in the cask. The casks will be designed to be transported on a 100-ton capacity railroad flatcar. The cask / car combination will be designed in accordance with DOT and NRC regulations, which include provision for crash protection and passive cooling capability. Casks designed to carry LWR spent fuel have been shown through test to provide substantial protecticn from credible, intentional destructive acts. A Exh 39 at 66, TR 3540; A Exh 35, Vol. 2, Chapter 5.7 at 5.7-50-S.7-51.
  --      - . , -  _ . _ .                      ,,      .r._ _-  ,
 
                              - F-114 -
than a million dollars per year. A Exh 39 at 64-65, TR I
3538-3539.                                                    l 138. The transportation of irradiated (spent) fuel and blanket assemblies removed from CRBRP represents a small incremental risk in addition to other fuel cycle operations. This risk is well recognized and DOE has substantial experience in shipping spent fuel from its various programs. A Exh 39 at 65, TR 3539; A Exh 35, Vol. 2, Chapter 5.7 at 5.7-50-5.7-51.
139. The spent fuel and blanket assemblies are hot, bcth radiolo-gically and thermally, and therefore, require special equip-ment for even the simplest handling operations. A Exh 35, Vol. 2, Chapter 5.7 at 5.7-50. It is undisputed that the material is highly unattractive as a target for the theft of plutonium, since chemical and mechanical operations requir-ing expensive complex facilities and equipment are needed to reduce it to a usable form. Moreover, without special shielding, radiation doses to individuals trying to work with unshielded or poorly shielded spent assemblies would be life threatening. A Exh 39 at 65, TR 3539; A Exh 35, Vol.
2, Chapter 5.7 at 5.7-50.
140. The safeguards cost of equipment and personnel for trans-porting CRBRP spent fuel will be about $200,000/yr. for the fourteen shipments needed per year. A Exh 39 at 67, TR 3541.
 
l
                                - F-ll5 -
CRBRP Waste l
141. There is no apparent dispute in the record concerning safe-i l      guards for wasto. Because of the low concentration of plutonium and uranium in radioactive wastes, low level A Exh 35, wastes are not considered attractive for theft.
Vol. 2,  Chapter 5.7 at 5.7-51; A Exh 39 at 77, TR 3551.
High level wastes do contain substantial radioactive material, and thus could be a target, albeit an unattractive one, for sabotage. A Exh 39 at 77, TR 3551.
142. High level radioactive waste (HLW) will be stored within the physical security bounds of the reprocessing plant prior to shipment. Due to the relatively high radioactivity and thermal output associated with HLW, transport to a reposi-tory will be accomplished in a fashion similar to the transportation of spent fuel. The high level waste will be shipped in a heavily shielded cask which will be resistant to penetration for sabotage. Safeguard requirements would be the same as those used for spent fuel.      A Exh 39 at 77-78, TR 3551-3552. See_ Findings 138-139 supra.
143. At the repository, the physical security of the site as well as the remote location of the wastes deep underground will effectively deter diversion or sabatoge. The requirements for protection against sabotage will be determined by NRC since this will be a licensed facility. A Exh 39 at 78, TR l
3552.
 
i
                              - F-ll6 -
144. The costs of adequate safeguards for waste are negligible.
A Exh 39 at 78, TR 3552.
145. The Department of Energy is bound by DOE Orders to provide      )
effeccive safeguards systems for the various fuel cycle        l facilities. A W Hammond, TR 3307-3309. Those systems will provide a level of protection against design basis threats which is at least equal to the level of protection provided by NRC requirements. See Finding 107, supra. Moreover, DOE is committed to providing an effective safeguards program for CRBRP fuel cycle facilities. A W Hammond, TR 3453-3455. In view of the safeguards requirements, current plans and designs for safeguards systems, the available tech-nology, and economic costs for safeguards, the radiological risks associated with safeguarding CRBRP fuel cycle facili-ties are small, and the economic costs are a small fractions (less than 2%) of total plant cost.
146. Intervenors did not submit testimony prepared by a safe-guards expert. Intervenors' Witness Cochran has never designed or assisted in the design of a safeguards system for a plutonium handling facility or a nuclear power reactor (I W Cochran, TR 3789-3790), has never reviewed a safeguards plan for a nuclear power plant (I W Cochran, TR 3790), has never conducted a physical inspection of a safeguards system for a nuclear power plant (I W Cochran, TR 3791), and is unfamiliar with, or has only limited familiarity with,
 
                                    - F-117 -
current safeguards technology (I W Cochran,' TR 3791-3794).
Intervenors' testimony did not dispute the fact that the CRBRP proposed safeguards system would meet all NRC require-ments.        Nor did Intervenorn dispute the fact that the pro-posed safeguards for the CRBRP fuel cycle will meet all DOE
                          */  Intervenors did, however, raise six allega-requirements.-
tions in support of their position that the NRC's analysis of safeguards is inadequate:              a) based on " empirical" evidence, thef t or sabotage is " credible", b) the Staf f fallad to analyze alternatives to the proposed reprocessing facility; c) possible civil liberties violations were not considered by the Staff, d) the Department of Energy will not implement an effective safeguards system at the fuel cycle facilities; e) MC&A systems must, but do not, indepen-dently deter theft or sabotage; and f) a clandestine fission explosive could be fabricated from 6 to 12 kilograms of plutonium using fresh LMFBR fuel without the need for conversion.
Empirical Evidence 147. Intervenors draw the conclusion that successful theft or sabotage is " credible" based on empirical evidence. IW Cochran, TR 3800; I Exh 12 at 12-13, TR 3899-3900.              The only
  --*/
At oral argument, counsel for Intervenors agreed that there were no essential factual disputes and that Intervenors' arguments were legal in nature. TR 6766.
 
l l
                                      - F-ll8 -
empirical evidence cited by Intervenors were:                        (1) "possi-ble" theft at the NUMEC plant; (2) "possible" theft of uranium at Wilmington, North Carolina; (3) sabotage of VEPCO Surry reactors; (4) sabotage of components for the Iraqi reactor                        i "while under fabrication in France"; and (5) Basque terrorists actions directed against Spanish nuclear facili-ties.        I W Cochran, TR 3800-3811; I Exh 12 at 12-13, TR 3899-3900.
148. As to the "possible" theft at the NUMEC plant, the theft occurred, if at all, in the early 1960's at a time when the safeguards and security requirements were virtually non-existent.        I W Cochran, TR 3800-3801.            In contrast, the NRC and DOE presently have in place extensive safeguards and security requirements which are fully applicable to CRBRP and its supporting fuel cycle facilities.                      Se'e 10 CFR Part 73; DOE Orders 5230, 5232; Finding 107 supra.
149. In regard to the "possible" theft of uranium at Wilmington, North Carolina, the facility in question handles only low enriched uranium -- not formula quantities of plutonium.
I W Cochran, TR 3801.          The safeguards requirements for facilities which handle formula quantities of plutonium, like CRBRP and its supporting fuel cycle facilities, are mere stringent than those for facilities handling only low enriched uranium.          I W Cochran , TR 3801.
l
 
                                - F-119 -
150. In regard to the sabotage of the VEPCO Surry reactors,      there was no release of radioactivity and in Dr. Cochran's opinion, the saboteurs' intent was only to cause property damage. I W Cochran, TR 3801-3803. In contrast, NRC's safeguards regulations are concerned with radiological sabo-tage. A Exh 39 at 27, TR 3501; 10 C.F.R. } 73.2(p). The fuel involved in the VEPCO Surry incident was low enriched uranium which Dr. Cochran presumed was not stored in a vital area. I W Cochran, TR 3804. In contrast, the fuel for the CRBRP will be stored in a vital area -- the reactor service building -- in molten sodium. I W Cochran, TR 3805-3806.
The security system at VEPCO Surry was not designed to meet current NRC requirements and was not required to meet NRC
                                                                  ~
requirements for strategic quantities of special nuclear material. I W Cochran, TR 3804. .NRC requirements for low enriched uranium are less stringent than those for strategic quantities of special nuclear material. I W Cochran, TR 3805. Dr. Cochran did not know where the low enriched uranium was stored at VEPCO Surry at the time of the sabo-tage, nor did he know the location of the fuel at CRBRP.      I W Cochran, TR 3805. Dr. Cochran was aware that CRBRP- fuel would be stored in sodium heated to 400-500 degrees F.      IW I
Cochran, TR 3806.
151. In regard to the " sabotage" of components for the Iraqi reactor while under fabrication in France, Dr. Cochran did
 
                                  - F-120 -
not know whether the facility was subject to safeguards, but he presumed it was not.      I W Cochran, TR 3807. Dr. Cochran knew nothing about the physical security at the French facility. I W Cochran, TR 3808.      He believed, however, that the physical security at the facility would not be as great as those imposed by NRC for facilities handling formula quantities of special nuclear material.        I W Cochran, TR r
3808.
152. In regard to the actions of Basque terrorists directed against Spanish nuclear facilities, Dr. Cochran admitted that the facilities were merely under construction -- not operating plants.      I W Cochran, TR 3808.
! 153. As noted in a GAO Report relied upon by Dr. Cochran, (I W Cochran, TR 3809, 3812) since January 1, 1977, there has been only one incident of an attack on a nuclear facility in the United States. This incident, however, involved placing an explosive device outside the plant gate, causing damage to a visitors' center.      I W Cochran, TR 3811, 3812. More-over, the NRC safeguards requirements for facilities dealing with formula quantities of SNM have been upgraded since j      January 1, 1977. I W Cochran, TR 3812-3813.
154. Citing an article in the Washi'ngton Post, Dr. Cochran also relied upon an " attack" at the Super-Phenix LMFBR facility in France. I Exh 12 at 15, TR 3902.      Dr. Cochran admitted
 
                                - F-121 -
that the facility was under construction -- not in opera-tion, that no radiological releases occurred and that the damage was to the side of the building with no breach of containment. I W Cochran, TR 3814-3816. Dr. Cochran had no firsthand information on the " attack" but relied solely on news reports such as the Washington Post for his informa-tion. I W Cochran, TR 3816.
Alternative Reprocessing Plants 155. As to Intervenors' contention that alternatives to the DRP were not considered, the uncontradicted record evidence demonstrates that the analysis of the DRP is, in fact, bounding. There are two alternatives to the DRP for repro-cessing CRBRP fuel. First, a small facility dedicated exclusively to CRBRP and FFTF fuels with approximately 15 f
A Exh 39 at 21, TR 3494. Second, a tons per year capacity.
breeder fuels head-end capability add-on to an existing LWR 1
l      fuels reprocessing plant. A Exh 39 at 21, TR 3494. The small dedicated Sacility can easily be safeguarded using conventional safeguards techniques. A Exh 39 at 21(a), TR 3495. Both the DRP and the head-end add-on to an existing high capacity LWR reprocessing plant would require more extensive safeguards systems and might utilize advanced techniques, particularly in the area of MC&A. A Exh 39      at 21(a) , TR 3495. The DRP, which has not yet been built,
 
                                  - F-122 -
represents the highest safeguards cost relative to the two alternatives.      See Findings 130-131 supra.
Civil Liberties 156. As to the possibility of civil liberties violations, includ-ing the imposition of martial law, the record evidence clearly establishes that such risks are no greater than those already encountered and accepted in the various mili-tary programs. I W Cochran, TR 3849. Moreover, in light of the fact that thef t or sabotage at CRBRP and its fuel cycle l
facilities is highly unlikely, (see Findings 106-145 supra) the possibility of any civil liberties violations is even 1ess likely. Finally, there is nothing in the record to suggest that such civil liberties violations have ever occurred either in the military programs or in the case of commercial reactors, or that the Board should presume that to be the case.
DOE Implementation 157. The record evidence in this proceeding clearly demonstrates that both the project and the DOE are strongly committed to I
implementation of effective safeguards systems for CRBRP and its fuel cycle facilities. A W Penico, TR 3450-3452, 3470; A W Hammond, TR 3452-3455. In addition, both the CRBRP and its fuel cycle facilities are subject to existing regulatory
 
                                  - F-123 -
requirements which mandate the implementation of an effec-l        tive safeguards system. A W Hammond, TR 3308-3309, 3452-3453. The CRBRP must meet the applicable saf eguards requirements of 10 C.F.R. Parts 70 and 73, while the fuel cycle facilities must meet the requirements of DOE Orders 5630 and 5632. A W Hammond, TR 3308-3309; S Exh 8 at E-9.
Applicants' commitments were reviewed by the NRC Staff and the Staff concluded that the commitments could be met and
                                                      */
would be met by the Applicants and DOE.--      S W Gaskin TR 3721-3722; S W Dube, TR 3683-3684, 3706; S Exh 8, App. E.
The Role of MC&A 158. Intervenors misconceive the nature and purpose of the physical security and MC&A systems within the safeguards system,. Thes'e systems are not intended by either DOE or NRC to be independently effective against threats of sabotage or theft. A W Hammond, TR 3363, 3432; S W Hurt, TR 3694-3695; S Exh 18 at 6, TR 3738; S W Dube, TR 3698-3699.      Rather, the two systems are complementary and are intended to and will,
    */  Contrary to Intervenors' assertions, the Staff conducted      a
    -~
systems review of all DOE fuel cycle facilities. S Exh 10 at 7, TR 3739. The review considered only existing tech-nology and gave no credit for DOE's research    and development efforts with the exception of NRTA. S W Dube, TR 3721. As to NRTA, the Staff concluded that the basic technology had been demonstrated, and could be implemented. S W Hurt, TR 3690-3691.
 
l                                                              - F-124 -
in combination, provide effective protection against theft or sabotage.        S W Dube, TR                  3698-3699. /
Clandestine' Fission Explosives 159. In their testimor.y, Intervenors claim that a clandestine fission explosive could be fabricated with 6 to 12 kilograms of plutonium using fresh LMFBR fuel without the need for conversion.        I Exh 12 at 7-8, TR 3894-3895; I W Cochran TR
                              **/      On cross-examination, Dr. Cochran conceded 3796-3797.---
that to obtain the six to twelve kilograms of plutonium from fresh mixed oxide fuel, it would be necessary to obtain at
    ~~*/
In contending that the MC&A system must be independently effective, Intervenors point to the fact that the limit of error on inventory difference (LEID) at the DRP would be greater than 2 kilograms of plutonium. Intervenors conclude from this alone that formula quantities of SNM could be stolen from the DRP without detection. I Exh 12 at 36, TR 3923. The record evidence clearly states that the MC&A systems and physical security system are complementary in nature and it is their combined effectiveness,which makes I
I theft or diversion of formula quantities of SNM a highly f                unlikely event.          See Finding 158 supra.
    **/          Intervenors assert that a plutonium dispersal device could be constructed using fresh mixed oxide fuel. I Exh 12 at 8, TR 3895.        In making this assertion, however, Intervenors overlook the fact that there are a number of radiological, chemical and biological agents which could be dispersed and which are not subject to extensive safeguards.                              Thus, plutonium in the CRBRP fuel cycle, which is subject to extensive safeguards is clearly an unattractive material for constructing a dispersal device.                          S Exh 10 at 9-10, TR 3741-3742; S W Hockert, TR 3663-3665.                          In addition, in light of the extensive safeguards associated with CRBRP and its fuel cycle facilities, diversion or theft for purposes of con-structing a CFE or dispersal device is a highly unlikely event. See Findings 106-145.
l l
 
                                - F-125 -
least three times the 6-12 kilograms because the fuel is only approximately 30 percent enriched. I W Cochran, TR I
3799. Dr. Cochran also conceded that each CRBRP fuel assem-bly weighs approximately 200-2S0 kilograms (approximately 450 pounds), and each assembly contains only 10 to 12 kilo-grams of plutonium. I W Cochran, TR 3826. Thus, in order to obtain 6-12 kilogram of plutonium from fresh CRBRP fuel, it would be necessary to steal three fuel assemblies, each of which is 14 feet long and weighs 450 pounds. A Exh 39 at 24, TR 3498. In addition, even assuming the highly unlikely theft of three fuel assemblies weighing 1350 pounds from CRBR, construction of a clandestine fission explosive (CFE) would be highly unlikely even for a person with a nuclear physics or nuclear engineering background, (S W Hockert, TR 3702-3703) and would depend to a large degree on luck. SW Hockert,  TR 3703. /
I l
i l
I
!    */  The CRBRP will not have weapons grade plutonium in its fuel. Weapons grade plutonium is defined as plutonium with less than 10 percent plutonium 240. S W Hockert, TR 3699.
CRBRP fuel will use FFTF grade fuel which has 12 percent plutonium 240. A W Strawbridge, TR 5164.
 
                                      - F- 126 -
l Contentions 6 b) 1) and 6 b) 3)'(Fuel ~ Cycle' Impacts) 160. Intervenors' contentions 6 b) 1) and 6 b) 3) allege the I
following:
l
!            6. The ER and FES do not include an adequate l
analysis of the environmental impact of the l
fuel cycle associated with the CRBR for the following reasons:
b) The analysis of fuel cycle impacts in the ER and FES are inadequate since:
T
: 1) The impact of reprocessing of spent fuel and plutonium separation required for the CRBR is inadequately assessed;
: 3)  The impact of disposal of wastes from the CRBR spent fuel is inadequately assessed;
                                                                                        */
Board Order of April 14, 1982, Appendix I at                                8-9.-
161. The CRBRP fuel cycle consists of four basic steps a) fuel fabrication for both core and blanket fuel, b) spent fuel reprocessing, c) associated waste management and d) transportation. S Exh 8, Appendix D; A Exh 35, Vol.                            2, l
l l
    */  Contentions 6(b)(1) and (3) were revised to delete the l
phrase "is not included or" preceding " inadequately assessed" l
to reflect the uncontested fa:t that environmental impact i        analyses for reprocessing of spent fuel and waste                                disposal Order Regarding are contained in the ER and FES Supplement.
Summary Disposition, October 26, 1982 at 4-7.                                Contention 6(b)(2), asserting that the impact of transportation of plutonium required for the CRBR is not included or is inade-quately assessed, was dismissed as moot.                                Intervenors so conceded and did not oppose the Staff's motion for and the Board's order granting summary disposition. Id. at 5-6.                                                  I Contention 6(b)(4), relating to the adequacy IBI safeguards in the CRBRP fuel cycle, is addressed in conjunction with Contention 4 supra.
i
 
                                  - F- 127 -
Chapter 5.7; A Exh 43 at 4, TR 4327. These steps are described in the following Findings.
162. Fabrication of mixed oxide (MOX) core fuel will be performed at the Secure Automated Fabrication (SAF) line, to be installed in the Fuels and Materials Examination Facility (FMEF), presently under construction at DOE's Hanford reservation. About 4 metric tons (MT) of the SAF line annual capacity of 6 MT MOX is needed to support the CRBRP l
fuel fabrication, or about 65 percent of the SAF line opera-tional schedule (15 of every 24 months). The basic SAF line l
fabrication process includes receiving and assaying of nuclear ceramic powders, blending of the powders, pelletiz-ing and sinterlag the powders into fuel pellets, and loading these pellets into finished fuel pins. S Exh 8 at D D-12; A Exh 35, Vol. 2,  Chapter 5.7 at 5.7.3 - 5.7.6; A Exh 43 at 6, TR 4329.
163. The CRBRP core and blanket fuel assemblies will be chemi-cally reprocessed to recover and purify uranium and plutonium for recycle back to the CRBRP.      Separation of the fission products from the fissile and fertile material is based upon liquid-liquid solvent extraction, utilizing the 1
l conventional Purex process, modified as required for speci-fic nuclear fuels. The ura.11um and plutonium products are l      converted to oxides in a form directly usable in fuel fabri-  )
cation. Although the specific facility for reprocessing I
l
 
                                    - F- 128 -
CRBRP fuel does not yet exist, Applicants' and Staff's analyses of fuel cycle impacts assumed that reprocessing of CRBRP (and other) fuels would be accomplished in The Developmental Reprocessing Plant (DRP).      S Exh 8 at D        D-17; A Exh 35, Vol. 2,  Chapter 5.7. at 5.7. 5.7-14; A Exh 43 at 7, TR 4330.
164. Low level wastes (LLW) produced during reactor operation and during fuel fabrication and fuel reprocessing operations will be transported to commercial shallow land burial sites for disposal. A W Newton, TR 4166. Transuranic wastes (TRU) produced during fuel fabrication operations at the SAF line will be stored at the existing DOE transuranic waste storage site on the Hanford Reservation and later shipped to l      a Federal repository for disposal.      TRU produced during operation of the DRP will be shipped to a Federal repository    l for disposal. High level waste (RLW) produced during fuel reprocessing at the DRP will be fixed in a matrix with a very low leach rate, packaged in stainless steel cylinders
                                                            */          i and shipped to a Federal repository for disposal.- A Exh        ;
l l
i
  --*/
Metal scrap generated at the DRP will be partially compacted, packaged into stainless steel cylinders and shipped to a Federal repository for disposal. Kr-85 and        ,
l I-129 are other wastes captured at the DRP. The Kr-85 (in a metal matrix) will be loaded into cylinders. One cylinder will be required for every 28 years of CRBRP operation. I-      )
129 will be fixed in concrete as barium iodate and packaged    '
in a 55-gallon drum. One drum will be required for every 20 years of CRBRP operation. Filled cylinders and drums will be sent to a Federal repository for disposal. S Exh 8 at D-(Centinued)
 
l l                                      - F- 129 -
43 at 7-8, TR 4330-4331; S Exh 8 at D-17-D-25; A Exh 35, Vol. 2, Chapter 5.7 at 5.7 5.7.21.
165. Operation of the CRBRP would require the transportation of radioactive materials between the power plant and the supporting fuel cycle facilities.      Commercial packaging and transport of radioactive materials are regulated at the Federal level by the Department of Transportation. Shipment by DOE is done in accordance with DOE Orders.        S Exh 8 at D-26-D-30; A Exh 35, Vol. 2, Chapter 5.7 at 5.7 5.7-39.
166. Applicants adequately analyzed the environmental impacts of each stage of the CRBRP fuel cycle and described the analyses and results in the Environmental Report (ER), A Exh 35, Vol. 2, Chapter 5.7. The Staff reviewed the Applicants' Jubmittals and performed an adequate independent assessment as tot  a) the reasonableness of the analytical approach, b) the credibility and conservatism of the assessment methods used by the Applicants, and c) the use of the best available information and analysis techniques.        S Exh 14 at 9-10, TR 4452-53.
167. The radiological impacts of the fuel cycle operations were the subject of dispute in the record. /        The pertinent l
20; A Exh 35, Vol. 2, Chapter 5.7 at 5.7-20-5.7-21; A Exh 43 at 8, TR 4331.
    -*/    The impacts of the CRBRP at each step of the fuel cycle were calculated by ascribing to the CRBRP a pro-rata share of the environmental impacts of a facility based on the percentage of a facility's capacity needed to support the CRBRP.        S Exh 8, Appendix D; A Exh 35, Vol. 2, Chapter  5.7.
 
                                  - F- 130 -
annual average population whole body exposures for the CRBRP fuel cycle are as follows:
CRBRP ' FUEL CYCLE RADIOLOGICAL' ENVIRONMENTAL ' IMPACTS */
Step          Annual Exposure (Person-Rem)
Core Fuel Fabrication                    0.1 Blanket Fuel Fabrication                0.1 Spent Fuel Reprocessing                  140 Waste Management                      small Total                              140 These exposures are small compared to the expected year 2010 U.S. population exposure due to natural background radiation of 28,000,000 person-rem. S Exh 8 at D-34; A Exh 43 at 9, TR    4332; S Exh 14 at 14, TR 4457.
168. During the course of the proceedings, Intervenors advanced f
the following arguments concerning the validity of the analyses of radiological impacts:    a) Applicants and Staff l
failed to consider the isotopic concentrations associated with the use of plutonium from recycled LWR or FFTF spent fuel in the dRBRP, and thus the dosos from reprocessing I      operations have been underestimated by a factor of 2 to 4.3.
(I Exh 13 at 25, TR 4591. ); b) Applicants and Staf f f afled to realistically estimate or conservatively bound the
  --*/
S Exh 8, Appendix D; A Exh 35, Vol. 2, Chapter 5.7; A Ex5 43 at 9, TR 4332. The non-radiological impacts as analyzed by the Applicants and Staff (S Exh 8 at D-8-D-9; A Exh 35, Vol.
'        2, Chapter 5.7 at 5.7-74-5.7-75) were undisputed.
 
                                      - F- 131 -
i environmental risks of CRBRP fuel reprocessing because they did not consider alternative reprocessing facilities to the f
DRP.    (I Exh 13 at 6, TR 4572.); c) Applicants and Staff failed to establish adequate containment factors for releases from CRBRP fuel cycle facilities. (I Exh 13 at 29-34, TR 4595-4600. ); and d) Applicants and Staff failed to properly evaluate the expected environmental impacts from radiological releases attributable to CRBRP wast manage-ment.    (I Exh 13 at 4-5, TR 4570-71. ) Each of these arguments is addressed in the following Findings.
Isotopic Concentrations 169. Intervenors contended that "by Staff's failure to consider plutonium from recycled LWR or FFTF spent fuel in the CRBR, Staff has underestimated the hazard of plutonium releases by a factor from 2 up to about 4.3."      I Exh 13 at 25, TR 4591. According to Intervenors, the use of spent LWR fuels with higher fuel burn-ups in the CRBRP fuel cycle would result in higher concentrations of the plutonium isotopes Pu-238 and Pu-241 than were assumed in the analysis, and i
therefore a higher dose attributable to plutonium releases than calculated by the Staff and Applicants. I Exh 13 at 19-25, TR 4585-91.
170. The Applicants intend to use fuel with a 12 percent Pu-240 content in the CRBRP. S W Lowenberg, TR 4380. The fuel cycle analysis done by Applicants was based on higher burnup
 
I
                                      - F- 132 -
l.
fuel (20 percent Pu-240). This is a conservative basis for l
assessing the reprocessing impacts since the facility would, in all likelihood, be processing lower burnup fuel (12
                            */ S W Lowenberg, TR 4380; S Exh 8 at D        percent Pu-240) .--
D-12.      The Staff used essentially the same composition of plutonium material assumed by Applicants for their environ-mental assessments of the fuel cycle facilities because the Staff " considered that composition to conservatively bound on the high side plutonium compositions that would be expected."      S W Lowenberg, TR 4383.
171. The fuel composition used by the Applicants in their fuel cycle analysis is equivalent to LWR fuel with a burn-up on the order of 20,000 megawatt days per metric ton. AW Yarbro, TR 4260. Since there exist ampic quantities of          .
available LWR spent fuel with burn-up less than or equal to that value, there would be no need to introduce higher burn-up fuel into the CRBRP fuel cycle during the five year
    */
The Staff calculated the radionuclide contentThe of CRBRP  spent Applicants f  --
fuel based on the use of 12% Pu-240 fuel.
used a more conservative set of values based 4380; on 20%IWPu-240 fuel. S Exh 8 at D-13; S W Lowenberg, TR Cochran, TR    4539. For assessing environmental effects from reprocessing, the NRC Staff used the higher of the two values of the source term for each individual isotope derived from Staff evaluations (NRC ORIGEN2 basis)  and Applicants' analyses (Section 5.7 of the ER). This app' roach thus uses the more conservative value for each isotope, resulting in an overestimate of environmental effects. A Exh 43 at 13, TR    4336; S Exh 8 at L -14.
l
 
                                                    - F- 133 -
                                                                                */
demonstration period.              A W Sherwood, TR 4313.--        Further-more,        the CRBRP license application is based on only the lower burn-up fuel and if the limits of the analysis were exceeded, the matter must be reviewed by the Staff.                      AW Strawbridge, TR 1751.
172. The major radiological environmental effects associated with the fuel cycle come from reprocessing. See table, Finding 167, supra. Furthermore, in reprocessing, the bulk of radiological impacts results from release of tritium and carbon-14. Over 99 percent of the dose to the total body of the U.S. population is due to those two elements. S Exh 14 at 22, TR          4465; S W Branagan, TR        4411; S W Clark, TR 4434.      Thus, even if it were assumed that the CRBRP were f                  fueled with higher burn-up plutonium fuel and that the plutonium source term in the two isotopes of concern, Pu-238 and Pu-241, were increased by a factor of 2 to 4.3, the result would be to increase plutonium doses due to plutonium releases by a factor of 2 to 4.3. Since these doses account for only one-tenth of one percent of the total
        */          Dr. Cochran admitted that little or none of existing spent
      -~
fuel from LWRs has higher burnup characteristics. IW Cochran, TR 4553.
          **/        Even if the fuel cycle started with spent LWR. fuel with a higher burn-up, as the fuel was recycled over time in the CRB RP , the concentrations of Pu-238 and Pu-241 would be reduced. A W Sherwood, TR 4265; A Exh 36, Vol. 3, Chapter 14 at 14.4A 14.4A-9; I W Cochran, TR 4539.
 
                                      - F- 134 -
dose, this is insignificant, A W Yarbro, TR 4265-4266, and would have no effect on the analysis.          S W Clark, TR
                  */
4434.-        Taking into account the conservatism of the analyses, the availability of ample lower burn-up LWR fuel, and the insignificance of the plutonium dose compared to the total dose from reprocessing operations, Intervenors' plutonium isotopics argument is not supported by the weight of the evidence in the record.
Alternative Reprocessing Facilities 173. Applicants' analysis was based on their present plan for carrying out reprocessing fuel at the projected Develop-mental Reprocessing Plant (DRP).      S Exh 8 at D-15.      The Staff independently evaluated the likely environmental
:            impact of the DRP drawing upon previous analyses of licensed reprocessing facilities, and other extant information on government facilities, NRC projections of radionuclide inventories, and plant separation factors.        S Exh 14 at 15, TR 44S8; S Exh 8 at D-12-D-17.
174. Two isotopes, tritium (H-3) and carbon-14 (C-14), dominate the estimated radiological impacts, providing over 99% of i
l
  */          Intervenors' testimony demonstrates that if bone-dose rather I
than whole-body dose were considered, the resultant effect due to plutonium would again be less than one percent of the total. I Exh 13 at 28, TR 4594. But cf. G Exh 8 at 12                12-64 (use of bone dose is inappropriate).
 
l I                                          - F- 135 -
                                                */  S Exh 14 at 22, TR 4465; A the estimated whole body dose.-
Exh 43 at 13, TR 4336.        The NRC Staff's source term conser-vatively assumed that all of the tritium produced at the reactor is transferred to and is released from the repro-cessing plant. In reality, however, about 90% of the tritium generated at the reactor will diffuse through the cladding into the sodium coolant, where it will be removed by the sodium cold traps.        Thus, the tritium source term and resulting doses are expected to be a factor of 10 less than the values calculated by the Staff in the FES.          A Exh 43 at 13, TR 4336.
175. The C-14 source term in the FES conservatively assumed that all the C-14 produced in both the fuel and the cladding is released during reprocessing.        In fact, the C-14 in the cladding remains with the cladding and would be disposed of at a permanent repository. As a result, the C-14 source term quoted in the NRC ORIGEN2 analysis is the more likely value and is a factor of 1.7 lower than the source term value used in the FES. In addition, C-14 that reaches the dissolver off-gas system will be removed along with the Kr-85 by the krypton removal system.          This is expected to
  */        This also .tolds true for bone dose.        See second footnote
  ~~
accompanying Finding 172, supra.
 
l l
1
                                    - F- 136 -
reduce the C-14 release by a factor of 2 to 10.        The com-bined effects of the corrected sourc,e term and C-14 reten-tion are expected to reduce the C-14 release and resulting environmental effects by at least a factor of 3 below that l      given in the FES.      A Exh 43 at 13-14, TR    4336-37.
176. The net effect of the Staff's assumptions is that the FES estimate of the U.S. total body population dose due to reprocessing is a factor of dbout 5 higher than the expected
                                                  */                            I doses and represents a bounding case.-        A Exh 43 at 15, TR 4338; S Exh 8 at D-12-D-17.
177. Intervenors contended that Staff and Applicants should have analyzed the environmental impacts of reprocessing at alter-native facilities to'the DRP, including existing facilities I Exh 13 at 6, TR 4572.      As at Savannah River and Hanford.
discussed in Findings 174-175, supra, however, the analysis of the environmental impacts of operation of the DRP assumed l
i I
While the source term estimates were bounding on the high-
  --*/    side, the Staf f's independent evaluation of fusi repro-cessing for CRBRP was based upon conservative (low side) assessments of the capabilities of the projected DRP to contain and retain the radionuclide effluents. This bound-ing assessment methodology assumes that the reprocessing activity for CRBRP accounts for about 80% of the radiologi-cal dose to the population from the entire CRBRP fuel cycle. However, despite this conservative approach, the Staff found that the radiological whole body exposure of the public from the entire CRBRP fuel cycle is very small (0.001%) compared with naturally occurring radioactivity.
This assessment is projected to bound the possible alter-natives for this activity and still results in small, essentially immeasurable, contributions to whole body population exposures. S Exh 14 at 18, TR 4461.
 
                                    - F- 137  -
that all of the tritium and all of the carbon-14 would be released to the environment, even though expected releases would be much lower. Such releases could not be exceeded no matter what reprocessing facility was used or where it was located. A W Sherwood and Yarbro, TR 4250-51. Therefore, the analysis of the DRP with a total release of H-3 and C-14 bounds all other potential reprocessing facilities. /      AW    ,
Sherwood and Yarbro, TR 4250-51; S W Lowenberg, TR 4405-06; A Exh 35, Vol. 2,    Chapter 5.7 at 5.7-14; S Exh 8 at D          D-17.
Containment Factors 178. Intervenors' Witness Cochran contended that the Applicants
                                                                        **/
and Staff have failed to use adequate containment factort--
for releases of plutonium from the CRBRP fuel fabrication
    --*/
The initial five years of CRBRP operation does not depend on completion of the Developmental Reprocessing Plant (DRP) prior to or during that time.      A W Hartman, TR 4317.
Furthermore, if reprocessing facilities were not available during the demonstration period, the spent fuel would have to be stored rather than reprocessed. The major portion of the overall fuel cycle environmental effect results from effluents from the reprocessing plant.      The environmental impacts of storing spent fuel are orders of magnitude less. Therefore, if the reprocessing facility were unavail-able during the demonstration period, the environmental effects of the fuel cycle would be markedly reduced from those contained in the FES. S W Lowenberg, TR 4439.
    **/  The containment factor, or clean-up factor, as used by Intervenors, is the ratio of the quantity of plutonium released to the environment divided by the total throughput of a facility during the same period of time. I W Cochran, TR 4541.
 
m .-      _      __ .
                                      '                                                  t        ,
g
                                                                                                      -[  Y',"
                                                                                                                        .t
                          ' '                                                                          .+
N,
                                                                              ' 71                            [
                                                          +-
                                                          .              - F- 138 -
j                                                    .
v i          and fuel reprocessing facilities.                                              I Exh 13 at 29-34, TR
;          4595-4600.                Intervenors advanced two specific babes _for
                                                                                                                              ~
I                                                                                -n  ?                      -      :<                  .
l          their contention:                                1)        that the bontainmentc;.              -
fact'or for- the I
CRBRP fuel reprocessing facility gill likely be a factor of l          ten greater than claimed by Staff and Applicants, based on operational experience at Hanford and Savannah River, I Exh 13 at 31-33, TR 4597-99; and 2) that the containment factors J
for the fuel reprocessing and fuel fabrication facilities failed to take into account operational experience at the LOE Rocky Flats facility, I Exh 13 at 33, TR                                                        4599.
j    179. Thu plutonium containment factors used by the Staff and
                                                                                                                                    ~
Applicants are 1.25X10                                          for fuel fabrication and 5X10 l            for reprocessing.                              S Exh 8 at D-lO-D-15; A Exh 35, Vol.-2, Chapter 5.7 at 5.7 5.7-25, I Exh 13 at 30, TR 4596.
The factors were based on the assumption that exhaust gases would pass through a series of high-efficiency particulate absolute (NEPA) filters, with each filter having an effi-ciency of at least 99.95%.                                        S Ebth 8 at D-11.                These values are consistent with the substantial base of experience in i
similar fuel cycle facilities and no reliable evidence in
                                                                                                                                */
the record contravenes the validity of these values. n i
      */      A substantial body of. knowledge exists regarding HEPA filter design Lt.d use.                    S W Lowenberg, TR 6084-85.                                  Filter perfor-mance has been sustained throughout accidents to about 99.9 percent efficiency or greater.                                        S W Clark, TR 4436.                        Fur-thermore, even if one stage of filtration were bypassed, the effectiveness of the remaining stages would not be impaired. S W Lowenberg, TR 4437.
 
l l
l
                                    - F- 139 -
l 180. Intervenors' Witness Cachran performed calculations based on l
      ' data from the Savannah River Plant and the PUREX Plant at Hanford and derived containment factors for those facilities 9          9 cf 4X10-  and 3X10    , respectively. These values are a factor of ten worse than that used by Staff and Applicants.
I Exh 13 at 31-33, TR 4597-99.      On cross-examination it was established that Intervenors' analysis is of questionable reliability and admittedly incomplete.                I W Cochran, TR 4565.
181. Even if Intervenors were scaehow correct in their view that Applicants and Staff overestimated containment capability by a factor of 10, effluent reductions of this magnitude can be achieved simply by the application of standard, well-proven engineering techniques.      Effluents can be reduced signifi-cantly by adding additional banks of HEPA filters.                SW Lowenberg, TR 4431-32; S Exh 8 at 12-61-12-62.              Addition of I        only one bank of filters would improve containment by a factor of 1000. Id. An improvement in containment by a factor of 10 could be achieved simply by increasing pipe or duct size, thus allowing material to more readily settle out S W Lowenberg and Clark, TR 4430-31..              Based of air streams.
on the evidence presented, the Board finds that the con-tainment factors assessed in the Staff's and the Applicants' analyses are reasonable and can be achieved.
l l
(                                            ,_    _ _ . _ .                __    _  _ _ _ _
 
                                  - F- 140 -
182. Intervenors' Witness Johnson argued that, based on experi-ence at the Rocky Flats facility, Applicants and Staff had underestimated the radiological releases from CRBRP fuel cycle facilities. Dr. Johnson's argument was premised upon
{        his assertion that the Rocky Flats facility is similar to the CRBRP fuel cycle facilities. I Exh. 21 at 2-4, TR 6019-21.
l l  183. Dr. Johnson's argument is not credible for a number of reasons. First, the CRBRP fuel cycle facilities are
,        dissimilar from Rocky Flats in terms of functions, products, and releases. Rocky Flats is not a fuel reprocessing or fuel fabrication plant, but rather is used for the fabri-cation and recovery of plutonium metal parts for weapons.
S W Lowcnberg, TR 6076. The incoming product in a nuclear fuel reprocessing plant, such as DRP, is spent fuel and the major radioactivity comes from the fission product content of the fuel. On the other hand, the main incoming products at the Rocky Flats plant are plutonium metal shapes, which essentially contain no fission products.      S W Lowenberg,
!        TR  6076-77.
  '.84. The releases from Rocky Flats and those from a fuel fabri-cation or fuel reprocessing facility likewise are also markedly different. For instance, there are little or no gaseous effluents in the form of fission products coming from the Rocky Flats plant. In the DRP the main releases of
 
                        -        -. ~
l                                - F- 141 -
l concern are the fission activation products such as C-14.
S W Lowenberg, TR  6077-78.-/    On the basis of the evidence in the record, the Rocky Flats facility is sufficiently dissimilar from the CRBRP fuel cycle facilities in terms of function, process, products and releases so that any attempt to use Rocky Flats as the basis for analysis of CRBRP fuel cycle impacts is meaningless.
  */  Mr. Lowenberg's discussion of the important differences
  ~~
between the Rocky Flats facility and CRBRP fuel cycle facilities was based on his considerable knowledge of and expertise in nuclear processing facilities, including Rocky Flats. S Exh 15, TR 4903; S W Lowenberg, TR 6075.
Dr. Johnson, on the other hand, had no formal training or experience which allowed him to compare the Rocky Flats f acility with CRBRP fuel cycle facilities.      I W Johnson, TR 5813-20, 5840-43.
  **/
Intervenors argue that the Staff's analysis of radiological releases from fuel reprocessing and fuel fabrication facili-ties was inadequate because it failed to consider the acci-dental releases, such as those from the fires at the Rocky Flats facility. I Exh 13 at 33-34, TR 4599- 4500; I Exh 21 at 4, TR 6021. Of  course, Rocky Flats is dissimilar from the proposed CRBRP fuel cycle facilities. Findings 183-184. Furthermore, subsequent to those fires, DOE formed a special task force to evaluate the occurrences, and imple-mented facility design changes which are now reflected in DOE orders, and new DOE facilities.      Staff Witness Lowenberg, TR 6079; I W Johnson, TR 5893-95. Applicants have committed to meeting applicable environmental release standards for the CRBRP fuel cycle facilities (A W Clark, TR 4390-91), including guidelines equivalent to the 10 C.F.R.
Part 100 for DRP. A Exh 35, Vol. 2, Chapter 5.7 at 5.7-8.
 
                                                                          - F- 142 -
Waste Management 185. The CRBRP fuel cycle facilities producing radioactive wastes l
(1) the blanket fuel fabrication plant, (2) the core
                                                                                          ~
are:
fuel fabrication plant, (3) the reactor plant, and (4) the fuel reprocessing plant.              A Exh 35, Vol. 2, Chapter 5.7 at 5.7-15-5.7-21.        For each of these facilities, the Staff independently assessed the quantity and types of radioactive i
waste that are likely to be generated over the life of the CRBRP . S Exh 14 at 16, TR          4459. The CRBRP wastes are similar to other wastes that might result from the commer-cial nuclear power industry and the portion of waste manage-ment facilities that might be required for CRBRP would be a                <
small fraction of the total waste management needs (i.e.,
less than 1%). S Exh 14 at 17, TR 4460.
186. The Staff estimated that 1% of a high level waste (HLW) i repository would be needed for CRBRP wastes, and based its astimate of CRBRP HLW management on that percentage.                S Exh 8 at D-25.      Since a more realistic estimate would be
                                                    */
: 0. 3 6 %  ,-- the environmental impacts of waste management
                                        */  The CRBRP wastes would occupy about 0.36% of the capacity of
                                      -~
a 2,000 acre repository.              The thermal design criterion cur-rently under consideration by DOE is 100 kw per acre of repository. Since the CRBRP waste produces about 4 kw per canister, the 180 canisters of CRBRP HLW will produce about 720 kw of heat and will require about 7.2 acres.              A Exh 43 at 16, TR 4338-39.
 
                                        - F- 143 -                            l contained in the FES (S Exh 8 at D-8-D-9), are conserva-tively overestimated by a factor of about 3.      A Exh 43 at 16, TR 4339.
187. Intervenors argue that the Staff incorrectly calculated the potential health effects associated with high level waste        l disposal. This argument is based on the Draft EPA Proposed l            Environmental Standards and Federal Radiation Protection l
Guidance for Management and Disposal of High-Level and l
Transuranic Radioactive Wastes, which contain limits based        l i
l                                                                              l on an upper bound value of 1000 health effects over the          j l
first 10,000 years after closure of a full-size high-level      f
                                                                              \
                                */
waste repository.-      I Exh 13 at 36, TR 4602.                  l 188. The health ef fects attributable to the CRBRP can be derived by taking 0.36-1 percent of the 1000 estimated health effects from the draft EPA standards. S W Boyle, TR 4422-4423. Intervenors, calculated and attributed 10 estimated health effects to CRBRP, but projected all 10 to occur within the thirty years of CRBRP operation, thereby yielding 0.3 health effects / year. I Exh 13 at 36, TR 4602. The health effects estimates contained in the EPA proposed standard, however, are upper level estimates of those. which might occur over a period of 10,000 years, not 30 years.
Thus, correct application of the EPA standard would show Intervenors concede that this is an upper bound value.      TR
        -*/    6939.
L
 
                                    - F- 144 -
that 0.00036-0.001 health effects per year could be expected from waste management activities attributable to CRBRP.-/ S l
I      W Boyle, TR 4422-23; A Exh 43 at 15-16, TR 4338-39.
189. The EPA proposed standards contain " pessimistic assumptions" f
which make the estimates contained therein the " upper bounds of the risks."      A Exh 44 at 9-10, I W Cochran, TR 4551-52, 6939. Consequently, contrary to Intervenors' arguments, the impacts of high level waste management activities attribut-able to CRBRP have been conservatively estimated ' and are not      -
significant.
i 1                                                                .
l l
l
    */    This is insignificant when compared to the background inci-dence of health effects. A Exh 42 at 24, 28, TR 4290, 4294.
 
                                                              - F-145 -
1 Contenti~ons ~ Sa l' and' 7c } "( Alternative" Sites')
190. Intervenors' contentions Sa) and 7c) allege the following: /
I          5. Neithe r Applicants nor Staff have established that the site selected for the CRBR provides adequate protection for public health and safety, the environment, national security, and national energy supplies; and an alterna-tive site would be preferable for the following reasons:
a)    The site meteorology and population density are less favorable than most sites used for LWRs.
(1) The wind speed and inversion conditions at the Clinch River site are less favorable than most sites used for light-water reactors.
l                        (2)  The population density of the CRBR site is less favorable than that of several alterna-
+                              tive sites.
(3)  Alternative sites with more favorable meteorology and population characteristics have not been adequately identified and analyzed by Applicants and Staff. The analysis of alternative sites in the ER and the Staff Site Suitability Report gave insuf-ficient weight to the meteorological and population disadvantages of the Clinch. River site and did not attempt to identify a site or sites with more favorable characteristics.
: 7. Neither Applicants nor Staff have adequately analyzed the alternatives to the CRBR for the following reasons:
        */  The issue of risk to nearby facilities resulting from acci-
        -~
dents at the CRBRP has been adequately addressed in the con-text of contention Sb), Findings 88 - 104, supra.                                  Since that risk is neither significant nor unacceptaofe for the CRBRP site, it is not a significant factor in the alterna-tive siting analysis.
 
                                  - F-146 -
c)  Alternative sites with more favorable environ-mental and safety features were not analyzed ade-quately and insufficient weight was given to environmental and safety values in site selection.
(1)  Alternatives which were inadequately analyzed include Hanford Reservation, Idaho Reserva-tion (INEL), Nevada Test Site, the TVA Hartsville and Yellow Creek sites, co-location with an LMFBR fuel reprocessing i
plant    (e.'g.', the Development Reprocessing Plant), an LMFBR fuel fabricating plant, and underground sites.
i Board Order of April 14, 1982, Appendix I at 6-7, 9-10.
191. In August, 1976, the Commission established certain prin-I ciples for the conduct of this proceeding and the review and
(
consideration of alternatives. These include: a) "the need I
for a liquid metal fast breeder reactor program, including I
its objectives, structure, and timing" shall be taken as established; b)      "the need for a demonstration-scale facility to test the feasibility of liquid metal fast breeder reactors when operated as part of the power genera-tion facilities of an electric utility system, including its timing and objectives" shall be taken as established; c) alternatives for meeting the objectives "are to be evaluated in terms of the objectives defined in the [ programmatic]
impact statement," and consideration of alternatives _will be l
limited to determining whether substantially better
 
l
                                                          - F-147 -
                                                                            */
alternatives are likely to be available.-              United States
!                      Energy ~Research''and*Devel'opment'' Administration (Clinch River l
Breeder Reactor Plant), CLI-76-13, 4 NRC 67, 92 (1976).
192. The programmatic objectives of the CRBRP are (1) to demon-strate the technical performance, reliability, maintaina-bility, safety, environmental acceptability, and economic f
feasibility of an LMFBR central station electric power plant i                        in a utility environment, and (2) to confirm the value of this concept for conserving important nonrenewable natural i
resources.            S Exh 7 at 1-1. The timing objective of the
{                        CRBRP is completion "as expeditiously as possible."            S Exh 8 i
at 8-4.
193. Following the Atomic Energy Commission's selection (from j
four alternative plant / site proposals submitted by utility
                */        The Commission has issued for comment a proposed rule for evaluation of alternative sites, 45 Fed. Reg. 24168 (April 9,          1980). S Exh 8, Appendix K. Although it is not con-trolling here, the methodology of the proposed rule has been i                          incorporated into both Applicants' and Staff's analyses as an additional confirmation of results. The. proposed rule              '
contemplates a two-part analysis: 1) comparison of environ-mental factors at the proposed site with those at alterna-tive sites to determine whether any alternative sites are
                          " environmentally preferred" to the proposed site; 2) if such a preferable alternative site exists,.a determination                '
whether that site is "obviously superior" to the proposed site based on a balancing of environmental and safety considerations, project economics, technology, and institu-
!                          tional factors. 45 Fed. Reg. 24177. "Obviously superior" and "substantially better" are regarded as essentially equivalent tests in Applicants' analyses. A W Kripps, TR 4693-4694. Intervenors agree that the terms are equivalent l
and that the test embodied in the proposed rule is not controlling.          I Counsel, TR 4659, 6953.
 
l
                                                                                        )
l I
                                  - F-148 -
I l
groups from across the nation) of the Tennessee Valley Authority / Commonwealth Edison Company proposal for con-l struction and operation of the LMFBR demonstration plant on the TVA system, the Clinch River site was selected as the A Exh 36, proposed site for the LMFBR demonstration plant.
Vol. 3, Chapter 9.2. In spite of the apparent practical need to locate the plant site in the TVA power service area, alternative siting analyses for the LMFBR demonstration plant have considered alternative sites not cnly within the TVA power service area, but also on all lands in the custody of DOE nationwide and on land in the custody of TVA outside of its power service area.      The alternative siting concepts of a "hcok-on" plant, of co-location with an LMFBR repro-cessing or fuel fabrication plant, and of underground siting have also been extensively examined.      A Exh 36, Vol. 3, Chapter 9.2; A Exh 37, Vol. 4, App. A,    D, E, F, and G; S Exh
                                                              */
7,  Chapter 9; S Exh 8, Chapter 9 and Appendix  L.--
    */    The nature and extent of alternative sites considered by Applicants far exceed those usually considered for LWR facilities. For instance, the Commission's proposed siting rule, footnote to Finding 191 supra, establishes the region of interest for alternative sites normally to be limited to the state in which the proposed facility is located or to the applicant's service area. The region of interest must be greater only if the environmental diversity would likely be substantially increased and if a) candidate sites within the initial geographic area meet the threshold criteria and the development of sites in the added geographic areas would likely not substantially increase costs or b) cy.ndidate sites within the geographic areas do not meet the threshold criteria and the development of sites in the added geo-(Continued)
 
l
                                                                                                                    \
                                            - F-149 -
f 194. The Applicants' alternative siting analyses, and the inde-                                                    :
l i
pendent review by the NRC Staff, showed that an all new                                                  !
!          plant at the Clinch River site is the preferred siting concept for the LMFBR demonstration plant, and accordingly that there are no substantially better siting alternatives for satisfying programmatic objectives.                              A Exh 45 at 4,              15-16, TR 4736, 4747-48; G Exh 16 at 16, TR 4922.
Alternative' ' Siting' Coricepts 195. Uncontested evidence supports the Applicants' and Staff's conclusions rejecting the alternative siting concepts of a hook-on plant, co-locatien, and underground siting. The hook-on concept would utilize turbine-generators at existing conventionally-fired electric generation plants to receive steam from the LMFBR demonstration plant nuclear steam supply system.        A Exh 36, Vol. 3, Chapter 9.2 at 9.2            9.2-8.        A stand-alone design is preferable, since potential dollar savings for the hook-on option cannot be realized, I
substantial schedular and economic penalties would result if the hook-on option were pursued, and the technological bene-fits of a stand-alone plant design are significantly greater than a hook-on plant design.                      S Exh 16 at 3, 6-7, TR 4909, 4912-13; S Exh 8 at 9-9; A Exh 36, Vol. 3, Chapter 9.2 at graphic areas would not require exorbitant costs.                                  S Exh 8 at K-9-K-10.
 
                                                            - F-150 -
9.2 9.2-33; A Exh 38, Vol. 5 Amendment XV at Q320.lR-
  .              1. Upon comparison of the safety, environmental accepta-bility, safeguards, and economic considerations, it was                            .
shown that the concept of co-locating the LMFBR demonstra-tion plant with an LMFBR reprocessing or fuel fabrication A Exh 37, plant is not a substantially better alternative.
Vol. 4, Appendix D at D D-96 and Appendix F at F-26; S Exh 7 at 11-37; S Exh 8 at 11-23; A Exh 45 at 11 TR 4743; 5                          '
Exh 15 at 31, TR 4895.                      Any expected safety benefits of underground siting do not offset the. penaltics (greater operational problems, major unresolved tuchnical prcblens, potential for groundwater contamination, significant increases in construction and operation costs, and a longer construction schedule) associated with such siting. Hence,                          ,
underground siting is not a substantially better alternative siting concept.                      S Exh 7 at 11-37, 11-38; S Exh 8 at 11                  11-24; A Exh 37, Vol. 4, Appendix D at D D-98 and Appendix F at F-26; A Exh 45 at 11, TR 4743; S Exh 15 at 28-29, TR 4892-93.
l Alternat'ive' Si~tes' Within' the 'TVA Power ~ ' Servi ~ce" Area 196. Af ter 109 possible alternative sites for the LMFBR demon-stration plant within the TVA power service area were screened, eleven sites were identified by Applicants as possible alternative sites.                    From these, the Clinch River site was selected as the preferred site.                      A Exh 36, Vol. 3, l
l
 
t f
!                                            - F-151 -
Chapter 9.2; A Exh 37, Vol. 4. Appendix A; S Exh 16 at 2, TR 4908. The Staff independently considered those eleven sites and correctly concluded that no significant environmental benefits would be gained by location at any of the alterna-
                    */ S Exh 7 at 9-3-9-10; S Exh 16 at 7, TR 4913.
tive sites.-
197. The Applicants performed a separate additional analysis of eleven candidate sites which are representative of the best sites within the TVA power service area.                      A Exh 37, Vol. 4, Appendix G, Attachment 1. These eleven sites were the                                      ,
proposed Clinch River site and ten alternative sites (Spring Creek, Blythe Ferry, Caney Creek, Taylor Bend, Buck Hollow, Phipps Bend, Lee Valley, Murphy Hill, Hartsville, and Yellow Creek). Comparison of the Clinch River site to the ten alternative candidate sites on the basis of hydrology, water terrestrial resources, quality, aquatic biological resources, water and land use, socio-economics, meteorology, and
  --*/
The Applicants, in 1982, updated the original alternative siting analysis using the approach set forth in NRC's Pro-posed Rule on Alternative Sites (45 Fed. Reg. 24168-24178, April 9,  1980). A Exh 37, Vol. 4, Appendix G. In con-for. nance with the proposal rule, the cpdated analysis demon-strated that a) the TVA power service area was an appropriate
        " region of interest", b) the TVA sites considered in the original analysis constitute a sufficient number of candi-date sites that meet the Proposed Rule's threshold criteria and represent the environmental diversity of the TVA power service area, and c) addition of applicable current informa-tion would not change the previous conclusions that the Clinch River site is the preferred location for the LMFBR Demonstration Plant.                A Exh 37, Vol. 4, Appendix G; A Exh 45 at 6, TR 4738.                                                    .
 
                                                                                                    - F-152 -
population showed that none of the ten alternative candidate sites was preferable to the Clinch River site from an environmental standpoint, and, therefore, that none is substantially better than the Clinch River site.                                                                  A Exh 45 at 7, TR 4739.
198. The NRC Staff independently reviewed the alternative sites, and selected the Hartsville, Murphy Hill, Phipps Bend and j
Yellow Creek sites along with the Clinch River site as an appropriate slate of alternative (candidate) TVA sites for S Exh 16 at 6, TR 4912. The NRC Staff correctly I                        analysis.
concluded that no alternative TVA site would be environmen-tally preferable (S Exh 16 at 13, TR 4919) and therefore none would be substantially better than the CRBRP site. S
                                                                                                  */
Exh 8 at 9-9.-
Consistent with the Endangered Species Act, the NRC asked
                    --*/    the U.S. Fish and Wildlife Service (FWS) to provide a cur-rent list of threatened or endangered species which might be affected by operation of the CRBRP. The FWS listed 1 species of bat, 1 species of fish and 11 species of fresh-water mussels. The Staff's analyses found no evidence of the bat species on the site, concluded that the project would not affect potential bat feeding habitats, and further concluded that the fish species das not present at the site.
A comprehensive TVA mussel survey of the Clinch River in the vicinity of the site collected no threatened or endangered species. However, during a sampling program for sauger the eggs, a single live speciman of Lampsilis''o. 'orbicul'at'a, the pink mucket pearly mussel, was found                                                    The in the vicinity Staff  completed        of an site.                                    S Exh 8 at 2-11, 2-19, 5-7.
endangered species assessment and submitted it                                                          to the Fish and Wildlife Service for approval in August                                                      1982. In that assessment the Staff concluded that construction and opera-tion of the CRBRP will not have an adverse ef fect on anyBy federally protected endangered or threatened species.
(Continued) l
 
i
                                                  - F-153 -
i Alternat'ive'TVA' Sites Outside^the'TVA' Power' Service ~ Area.
199. Uncontested evidence supports Staff's and Applicants' analyses rejecting alternative TVA sites outside the TVA power service area. Only two TVA sites outside the TVA service area, Page and Artemis, might reasonably be considered as alternative sites. Both, however, are unsuitable by reason of the inadequate size of the Page site, and limited water supplies, absence of barge access, 1
and high trancaission line costs at both sites.          5 Exh 7 at 9-11; S Exh 8 at 9-11    A Exh 37, Vol. 4, Appendix D at D                                      -
D-93 and Appendix F at F-26.
DOE' Sites I                200. In 1976, the Applicants surveyed all properties in the custody of ERDA throughout the United States in order to identify potential alternative sites for a LMFBR demonstra-i                        tion plant. After consideration of site size, availability {{letter dated|date=September 17, 1982|text=letter dated September 17, 1982}}, FWSSadvised        the NRC that it Exh 8 at 5-7, B                          concurred in the Staff conclusions.
B-8.
A question arose on the potential impact of the plant's thermal discharge on striped bass utilizing theThe    Clinch        River striped as a late summer, early fall thermal refuge.
bass are a game fish introduced into Tennessee watersBased          by way    on of a state stocking program. A W Kripps, TR 4719.
analysis to date, no adverse impactHowever, on the striped TVA has an on-bass is anticipated. S W Leech, TR 4770.
going program to evaluate the impact of thermal plumes on striped bass, and if significant impacts are projected, Applicants would be required to take appropriate remedial action. S W Leech, TR 4771-4772.
l
 
l
                                - F-154 -
of cooling water, sei smic ground motion, potential interfer-ence with other programs (e.g., those under the Division of Military Applications weapons program), population density, space for plant location, and proximity to existing ERDA facilities, the Hanford Reservation, the Idaho National Engineering Laboratory (INEL), and the Savannah River Plant (SRP) were identified as potential alternative (candidate) sites for the LMFBR demonstration plant.      S Exh 7 at 9-11; A Exh 37, Vol. 4, Appendix D. The most recent updated information showed that Hanford, INEL, and Savannah River still remained potential DOE alternative (candidate) sites for siting of a LMFBR demonstration plant. A Exh 37, Vol.
3 4, Appendix F at F-4; S Exh 8 at 9-11; S Exh 16 at 8-9, TR 4914-4915. /
201. Exanination of the environmental and engineering character-istics of the Hanford, Savannah River, and INEL sites demon-strated that none of those sites are environmentally prefer-able to the Clinch River site. S Exh 8 at 9-11; A. Exh 37, Vol. 4, Appendix F at F-8. While these sites have somewhat
  */  The Nevada Test Site was not considered suitable because of the estimated 0.75g design requirement for seismic ground I      motion, lack of surface water and limited groundwater (use l        for the demonstration plant would conflict with other uses of Nevada's limited supply) and relatively high transmission line costs. Potential interference with activities asso-ciated with research, development, and testing of nuclear weapons was also indicated. S Exh 8 at 9-11; S Exh 16 at 10, TR 4916; A Exh 37, Vol. 4, Appendix D at D-21-D-24.
 
                                                                          - F-155 -
i more favorable atmospheric dispersion (meteorology) and site isolation (i.e., minimum exclusion boundary distance, sur-rounding population density) characteristics than the Clinch River site (discussed i'nfra), the comparison of all relevant siting parameters showed the Hanford, Savannah River, and INEL sites are essentially equivalent to the Clinch River site in terms of environmental considerations and none are substantially better from an environmental standpoint. A Exh 45 at 9-10, 15, TR 4741-4742, 4747.
Intervenors' Arguments'.
202. The record shows that from the standpoint of environmental considerations alone, none of the TVA or DOE sites are preferable, much less substantially better, than the Clinch River site. Findings 196-201, supra.                              Intervenors argue, however, that the Hartsville and Yellow Creek TVA sites, and I                        the Hanford, Savannah River, and INEL DOE sites exhibit more favorable meteorological and population density characteris-tics, (and hence lower radiological risk), and for those reasons, are substantially better alternatives. TR 6953-
: 74. In the following Findings these arguments will be addressed with respect to the elements of meteorology, popu-lation density and radiological risk.                                    In addition, the elements of cost and programmatic factors, and the balance of all relevant factors will be addressed.
e
 
I
                                                        - F-156 -
Meteorology 203. Although certain alternative TVA sites have slightly better atmospheric diffusion characteristics (A Exh 37, Vol. 4, Appendix G, Attachment I), the atmospheric diffusion charac-teristics for the Clinch River site and for the alternative TVA sites can be considered comparable.                      S Exh 15 at 14, TR 4878; S W Spickler, TR 4805, 4811.                      The diffusion charac-teristics of tha three DOE alternative sites were found to be somewhat more favorable than those at the Clinch River S Exh 8 at L-34, L-40, L-44 L-44; S Exh 15 at 15, TR site.
4879.            A Ixh 37, Vol. 4, App. F at F-5; A Exh 45 at 9, TR i
4741.            The Clinch River site, however, is an acceptable site for a nuclear facility from the standpoint of meteorology.
S Exh 1 at IV-1-IV-2; A Exh 37, Vol. 4, Appendix D at D-2,
                                                                                                  */
D-9 and Appendix F at F-5; A Exh 45 at 14, TR 4746.--
Popul'at' ion' Densi ty I,
l 204. The O to 30 mile population density of the Clinch River site 1
l                      is average when compared to LWR sites.                      S Exh 15 at 22, TR 4886.            Although Regulatory Guide 4.7 states that areas with
      --*/
The meteorological data for the Clinch River site were collected by Applicants in accordance with Regulatory Guide 1.23.          S Exh 15 at 8, TR 4872. These data were then used to calculate the atmospheric diffusion factor (X/Q) in accord-ance with Regulatory Guides 1.111 and 1.145. S Exh 15 at 9-10, TR 4873-74. The Clinch River atmospheric diffusion conditions were then compared to those of alternative candi-date sites. S Exh 8, Appendix                L; S Exh 15 at 9-14, TR 4G13-78; A Exh 45,at 12, TR 4744.
l l
 
                                    - F-157 -
low population densities are to be preferred for the siting of nuclear power reactors, it does not draw any distinction as between siteP With " low popdlation densities" below certain " trip levels."    S Exh 15 at 23, TR 4887. The Clinch River site and the alternative TVA and DOE sites all have population densities below the trip levels. A Exh 45 at 13, TR 4745. Consequently, any differences in population density as may exist between Clinch River and the alterna-tive sites are insignificant, and no alternative site is substantially better than the Clinch River site with regard to population density. S Exh 15 at 22-23, TR 4886-87.
Radiologi~ cal' Risk .'
205. Sites with more favorable meteorology and population density,
!            all other factors being equal, would show lower offsite l
doses in the event of releases of radioactivity. A Exh 45 at 14, TR 4746. Intervenors argued that offsite doses would be less at the alternative sites by factors ranging from three up to twenty-four. TR 6956. These consequences, standing alone, are not important. Rather, it is necessary to consider whether the meteorology and population density differences between Clinch River and the alternative.-sites give rise to significant differences in risk.      In this regard, the record shows that: a) the doses at the Clinch River site meet the site suitability dose guidelines for any
 
                                - F-158 -
design basis accident (S W Soffer, TR 4818; S Exh 1 at III-10; A Exh 1 at 47-52; Findings 23 - 24, supra); b) the effects of routine releases at the Clinch River site are already so small that a further reduction would not consti-tute a significant reduction in risk (A Exh 45 at 14-15, TR 4746-47;    Findings 224-258, infra); and c) the risk of severe accidents beyond the design basis is already low and a further reduction would not constitute a significant reduction in risk.    ( A Exh 45 at 14-15, TR 4746-47; S W Sof fer, TR 4789; S Exh 8, Appendix J; Findings 22,73 - 87, supra). Accordingly, from the standpoint of radiological risk, the alternative sites are not substantially better than the Clinch River site.
Cost'and' Programmatic' Factors 206. Relocation to another TVA site would result in increased costs to the project of $39-303 million on a 1982 present worth basis and considerably more on an appropriations basis. Relocation costs to a DOE site, on a present-worth basis, are $94 million for relocation to Hanford, $259 million for relocation to INEL, and $61 million for reloca-tion to Savannah River. S Exh 8 at 9-12-9-14; S Exh 16 at 16, TR 4922; A Exh 37, Vol. 4, Appendix F at F-31 and Appendix G at G-28.
 
                                - F-159 -
207. If any alternative site were selected for relocation,    a minimum delay of 33 months and a more probable delay of 43 months could be expected, starting from the time a decision 1
was made to change sites. A Exh 37, Vol. 4, Appendix E at E-ll-E-19, Appendix F at F-28, Appendix G at G-25-G-26; S Exh 8 at 9-12; S Exh 16 at 14, TR 4920.      Unless offsetting benefits were present, delays ranging from 33 to 43 months would not be consistent with DOE's timing objective under l
l the LMFBR program - i.e. , completion of the project as        ,
expeditiously as possible. A Exh 37, Vol. 4, Appendix F st F-28, Appendix G at G-26; A Exh 45 at 8, 10, TR 4740, 4742; i
l      S Exh 16 at 15, TR 4921.
208. The project objective of utility participation and demon-stration in a utility environment is not likely to be met at the alternative DOE sites. A Exh 37, Vol. 4, Appendix F at F-8, F-30; S Exh 16 at 15, TR 4921.
209. From the standpoint of cost and programmatic considerations, Clinch River is substantially better than the alternative l      TVA and DOE sites. On' balance of all relevant environmental considerations (including meteorology, population density, and radiological risk) and cost and programmatic considera-tions, none of the alternative DOE and TVA sites is a sub-stantially better alternative than the Clinch River site for meeting the LMFBR program objectives.
 
                                                                          - F-160 -
Contentions 7 a) and b) (Likelihood ~of Meeting Program Objectives and Design Alternatives Intervenors' contentions 7 a) and b) allege the following :--
                                                                                                  */
210.
: 7.                              Neither Applicants nor Staff have adequately analyzed the alternatives to the CRBR for the following reasons:
a)  Neither Applicant 7 nor Staff have ade-quately demonstraced that the CRBR as now planned will achieve the objectives established for it in the LMFBR Program Impact Statement and Supplement.
(1)  It has not been established how the CRBR will achieve the objectives there listed in a timely fashion.
(2)  In order to do this it must be shown that the specific design of the CRBR, particularly core design and engineering safety features, is sufficiently similar to a practical commercial size LMPBR that building and operating the CRBR will demon-strate anything relevant with respect to an economic, reliable and licensable LMFBR.
(3)  The CRBR is not reasonably likely to demonstrate the reliability, maintainability, economic feasibil-ity, technical performance, envi-ronmental acceptability or safety of a relevant commercial LMFBR central station electric plant.
b)  No adequate analysis has been made by Applicants or Staff to determine whether the informational requirements of the LMFBR program or of a demonstration-scale facility might be substantially better satisfied by alternative design
          */        Board Order of April 14, 1982, Appendix I at 9-10.
 
l l
                                            - F-161 -
features such as are embodied in certain foreign breeder reactors.
211. Intervenors' Contentions 7a) and 7b) raise three basic issues:        a)      whether the CRBRP is reasonably likely to meet its programmatic objectives (Contention 7a)(1) and (3));
b) whether the CRBRP will provide information relevant to commercial size LMFBRs (Contention 7a)(2)); and c)              whether the informational objectives of the CRBRP might be sub-stantially better satisfied by alternative design features different f rom those in the CRBRP.              (Contention 7b)). The evidence introduced by Applicants and Staff with respect to each of those issues is essentially uncontested.
212. The LMFBR program objectives and timing for CRBRP are set forth in the DOE Supplement to the LMFBR Program Final Environmental Impact Statement (FEIS), DOE /EIS-0085-D (May 1982) on page 57 as follows:
                    . to demonstrate the technical perfor-
  '                      mance, reliability, maintainability, safety, environmental acceptability, and economic feasibility of an LMFBR central station electric power plant in a utility environment;
                    . to confirm the value of this concept for
<                        conserving important nonrenewable                  -
natural resources.                              ,
S Exh 7 at 8-4; S Exh 8 at 8-2; A Exh 34, Vol. 1, Chapter 1 at 1.2-1.      In addition, the programmatic timing of the CRBRP has been established by the DOE FEIS and its record of decision to be "as soon as possible."            See 47 Fed. Reg.
 
                                      - F-162 -
33771 (August 14, 1982) . A Exh 58 at 4, TR 6410; S Exh 21 at 2, TR 6523.
' Achievement'of' Objectives 213. The objectives of the project have been made an integral part of the design and management process for the CRBRP.
i The project objectives were incorporated in all levels of the design through descending tiers of increasingly more detailed and specific design guidelines, the Overall Plant Design Description, the System Design Descriptions for each of the fif ty-six plant systems, and the Equipment Specifica-tions for each component in the plant.          A Exh 58 at 4-6, TR 6410-12. The project has adopted a series of formal manage-ment systems, including design reviews, configuration management, and quality assurance, to assure that the ability of the design to meet its objectives is controlled and measured on a continuous basis.          A Exh 58 at 6-14, TR 6412-20. Achievement of each specific project objective will be substantially assured by this systematic, l
disciplined management approach.
j 214. In regard to technical performance, CRBRP is likely to achieve its major technical parameters:        a) plant thermal l
power production, b) steam conditions, and c) electrical power production. A Exh 58 at 14, TR 6420; S Exh 8 at 8          8-4; S Exh 7 at 8-5.
 
                                          - F-163 -
215. Thermal power' production is a function of core heat genera-tion, core flow, and heat transport from the core in the heat transport system.      Planned core heat generation is likely to be achieved, based upon analysis of CRBRP core physics and comparison of these analyses with experiments conducted at the Zero Power Plutonium Reactor (ZPPR) using a CRBRP core configuration mock-up.      A Exh 58 at 14-15, TR 6420-21. Core flow characteristics have been determined by scale-model hydraulic tests, and the analytical tools for calculating basic heat transfer from the core are well established through experience with the Experimental Breeder i            Reactor-II (EBR II), the Fast Flux Test Facility (FFTF), and light water reactors (LWRs).      The overall heat transport system (HTS) is likely to meet the design parameters for plant thermal power production based upon experience from EBR II and FFTF.-/      A Exh 58 at 14-15, TR 6420-21.
l 216. The CRBRP steam conditions of importance are pressure, temperature, and flow. The steam, feedwater and condensate systems for CRBRP are similar to those currently in use in LWR's and fossil power plants, and the CRBRP conditions of
        --*/
The major HTS components are sodium pumps and Intermediate Heat Exchanger. A prototype of the main sodium pump is currently being tested and has been found to perform satisfactorily to date.      The Intermediate Heat Exchanger (IHX) is similar to the one successfully used in FFTF and l              can be reasonably expected to perform acceptably in CRBRP.
A Exh 58 at 15, TR 6421.
 
1
                                      - F-164 -
pressure, temperature and-flow fall within the range of parameters experienced for LWR's and fossil-fueled plants.
A Exh 58 at 15-16, TR 6421-22.
217. Intervenors have argued that high technical risks ar.d long project delays could occur if a steam generator of untested design were installed in the CRBRP. The steam generator design and verification tect program are well founded and based upon more than 20 years of relevant experience. AW Longenecker, TR 6325. The CRBRP steam generator design incorporates lessons learned from operating LMPBR steam generators as well as from LWR steam generator experience.
S Exh 21 at 8, TR 6529.' Model steam generators have been tested to obtain data on full-power steam generator per-formance and endurance. I Exh 22, Attachment, TR 6253. A prototype steam generator has been developed to perform component and system integration testing.  .S Exh 21 at 9, TR i
6530. Hydraulic testing of a 0.42-size scale model is planned in the future to confirm analytical predictions that i
there will be no flow-induced vibration problems with steam generator design improvements. /  Finally, as a confirmation
      --*/  The hydraulic testing on the 0.42-scale model and the plant spare unit should disclose any unforeseen flow-induced vibration problems. Thereafter, steam generator problems which might occur would be corrected in place, probably by plant operations personnel and designers working together.
An alternative course requiring fabrication and testing of a precise prototype before contracting for production of the (Continued)
 
                                                              - F-165 -
l                                  test of the scale model tests, the plant spare steam genera-f l                                  tor will be hydraulically tested.                S Exh 21 at 10, TR 6531.
218. The only evidence presented questioning the CRBRP steam generator design was a GAO letter entitled " Revising the
;                                  Clinch River Breeder Reactor Steam Generating Testing Program Can Reduce Risk," GAO/EMD-82-75, May 25, 1982, Attachment to I Exh 22, TR 6250-60.                In that letter, how-ever, the GAO acknowledged:                1) that all steam generator problems are not related to design deficiencies, 2) that
                                                      ~
testing cannot eliminate all elements of risk, and 3) that the ultimate test must come when the steam generators-are operated in CRBRP. Id. at 9, TR 6258.              The Staff and the GAO's technical consultant agree with the Applicants that the assurances gained from testing a precise prototype prior to manufacturing the plant units cannot technically justify the delay required to construct such a prctotype.                              I Exh 22, Attachment 2 at 9, TR 62S8; S Exh 21 at 10, TR 6531. The technical risk of an undetected major design defect, requir-ing redesign and lengthy delay after installation, is very small. S Exh 21 at 5-11, TR 6526-32; A W Longenecker, TR 6328.
i l
 
                                          ->F-166 -
219. Electrical'' power' product' ion for CRBRP will be achieved through the use of a turbine generator which will operate at conditions of temperature, pressure, and flow which fall l
within the range of parameters experienced for LWR's and fossil plants. A Exh 59 at 16, TR 6422.        On the basis of the foregoing and Findings 214 -        219 aupra, CRBRP is reasonably likely to meet its technical performance ebjectives.
220. The CRBRP has been designed to reach a baseload reliability of about 75% within the 5-year demonstration period.        The Applicants have made formal reliability analyses an integral part of the design process from the outset.        The availa-i bility of each CPBRP component and subsyst'em has been assessed using an existing data base containing the avail-ability performance data of similar componento and systems to assure a high likelihood that the plant would meet its reliability goal.-
                                */ S Exh 8 at 8-4; A Exh 34, Vol. 1,
  */      The assessment identified specific elements of the design which could be improved so that the Project's ability to meet the goal would be enhanced.        Two examples illustrate this point.      First, as a result of this review, it was determined that providing redundant heaters on the equali-I zation lines would increase the plant availability by j          1.6%. Secondly, inclusion of certain piping changes would l
allow maintenance operations for one Radioactive Argon l
Processing System (RAPS) compressor while the other is operating, thereby increasing the plant availability by 0.6%. Action has been taken to implement both of'these design modifications for the CRBRP.        A Exh 58 at 17-18, TR 6423-24.
 
                                                  - F-167 -
Chapter 1 at 1.3-2; A Exh 58 at 17-18, TR 6423-24; S Exh 21 at 15-17, TR 6536-38.
l 221. The CRBRP design includes specific features and requirements to enhance maintainabilityi-
                                                    */  Maintainability reviews are required parts of the design and design review process, (A Exh 58 at 18, TR 6424) and sound maintainability require-ments have been implemented in the design including:      a) all in-sodium components must be designed to drain freely of sodium so that, upon removal, liquid sodium does not freeze inside the components and thus complicate maintenance operations; b) major components must be either removable or repairable in place; and c) ample space must be provided around all major equipment to assure ease of access for
                                    **/                            It is likely that maint enance .--    A Exh 58 at 19, TR 6425.                      ,
                --*/
Maintainability encompasses the ability of the plant operator to perform preventive and corrective maintenance on the plent with minimal cdverse impact on the amount of time the plant is available for generation of electricity.      The actual maintainability goals to be achieved are constrained by requiring maintenance in a time frame that supports the plant reliability goal. A Exh 58 at 18, TR 6424.
                **/  In order to assure that this requirement would be met, the
                -~~
Applicants developed a detailed scale-model of the Clinch River Breeder Reactor Plant (one-half inch to one foot).
This scale-model has been applied as an engineering tool in review of all equipment arrangements to assure that no unforeseen interference would occur which could impact main-tainability. A Exh 58 at 19, TR 6425. In specific areas of the design where maintenance operations are expected to be critical to meeting the availability objectives, detailed models were built to verify that maintenance operations could be performed satisfactorily. A Exh 58 at 21, 23, TR (Continued)
 
1
                                                    - F-168 -
the CRBRP will meet its maintainability objectives.                              S Exh 7 at 8-6; S Exh 8 at 8-2; A Exh 34, Vol. 1, Chapter 1.0 at 1.3-3; A Exh 58 at 18-23, TR 6424-29; S Exh 21 at 17, TR 6538.
i 222. In regard to the safety objective, the NRC Staff's June 1982 Site Suitability Report concluded that "...the proposed CRBRP site is suitable for a facility of the general size and type proposed from the standpoint of radiological health and safety considerations." S Exh 1 at I-4. In addition, the NRC Staff's February 1977 Final Environmental Statement concluded that "...it is within the state-of-the-art to design, construct and operate the CRBRP in such a manner that the consequences of accidents will not be significantly different from those already assessed for LWRs."                              S Exh 7 at 7-11.                  The NRC Staff's Final Supplement to the Final Environmental Statement confirms this conclusion. S Exh 8 at 7-3. The CRBRP is likely to meet its safety objective. S 6427, 6429.                  For example, because the reactor head access area has both a relatively high density of equipment and the regairement for equipment movements during refueling opera-tions, a full-scale mock-up of the reactor head access area was constructed and used by the reactor component and systems designers to ensure that necessary operations and mainte-nance activities could be accomplished in the reactor head access area. As a second example, the high density of equipment in the area surrounding the reactor head made it necessary to construct a full-scale mock-up of the secondary control rod drive mechanism so that the designers could simulate and fully characterize the actual maintenance operations anticipated for those components. A Exh 58 at 21, 23, TR 6427, 6429.
 
                                - F-169 -
Exh 7 at 8-6; S Exh 8 at 8-4; A Exh 34, Vol. 1, Chapter 1.0 at 1.3-3; A Exh 58 at 23-24, TR 6429-30; S Exh 21 at 22-23, TR 6543-44.
223. In regard to environmental ~ acceptability, the CRBRP will satisfy all applicable Federal and State environmental regulations. A Exh 36. Vol. 3, Chapter 12.0. The NRC Staf f's Final Environmental Statement concluded that the environmental impacts of construction and operation were acceptable. S Exh 8 at v. It is likely that the CRBRP will meet the objective of environmental acceptability. A Exh 58 at 24, TR 6430; S Exh 7 at 8-7; S Exh 8 at 8-2; A Exh 34, Vol. 1, Chapter 1.0 at 1.3-4.
224. The economic feasibility objective will be achieved by developing comprehensive cost, material quantities, and performance information for the CRBRP for use in developing commercial-size central station power plants. The project has established a system for compiling this comprehensive cost information in a form which permits cost analysis and evaluation for all the plant elsments at a very detailed level. These CRBRP data have already been used in develop-ment of the LDP cost estimate, and in the future, the cost and performance data established for the CRBRP will be used to extrapolate the cost and economics of future commercial LMFBR plants. The CRBRP is reasonably likely to meet the
 
                                  - F-170 -
                                                          */
objective of demonstrating economic feasibility.-      A Exh 58 at 24-25, TR 6430-31; S Exh 21 at 18, TR 6539; S Exh 7 at 8-7; S Exh 8 at 8-2; A Exh 34, Vol. 1, Chapter 1.0 at 1.3-5.
225. The CRBRP will achieve its objective of operating in a utility' environment by operation on the Tennessee Valley Authority (TVA) system, supplying power to that grid, while being operated by TVA personnel. A Exh 58 at 25, TR 6431; S Exh 7 at 8-7; S Exh 8 at 8-2; A Exh 34, Vol. 1, Chapter 1.0 at 1.3-5.
226. In regard to the objective of confirming the value of the LMFBR in conserving important non-renewable natural resources,    the CRBRP will generate electricity utilizing an
                                                            **/
The otherwise unuseable natural resource--uranium-238.---
CRBRP is likely to meet the cbjective of confirming the value of the LMPBR concept for conserving important non-renewable resources. A Exh 58 at 25-26, TR 6431-32; S Exh 7
  */    While demonstration of a breeding gain is part of the plant's
-~
design guidelines, the attainment of a specific threshold value of breeding ratio is not, in the overall context of the LMFBR program, a priority for the CRBRP. A W Longenecker, TR 6382-83; See Board Order, dated October 5, 1976 at 7-10.
See also UnTted States Energy Research and Development
        %3 ministration (Clinch River Breeder Reactor Plant), CLI          13, 4 NRC 67 (1976). A respectable breeding ratio of at least 1.2 can be achieved with the current heterogeneous core design, and as plant scaleup increases in future designs, this core concept will make even higher breeding A W Anderson, TR 6383-85.
ratios readily achievable.
    **/ See footnote accompanying Finding 224, supra.
 
l
                                      - F-171 -
at 8-8; S Exh 8 at 8-2; A Exh 34, Vol. 1, Chapter 1.0 at 1.3-6.
227. The programmatic timing of CRBRP contemplates completion of CRBRP as soon as possible.      Project research and development and the design are almost complete.      Most of the hardware is on order or delivered and site preparation activities have commenced. The NRC Staff has issued a favorable Site Suitability Report (S Exh 1) and Final Environmental Statement and Supplement (S Exh 7 and 8) for the project.      It is likely that CRBRP will meet its objectives in a timely manner.-/        A Exh 58 at 26-27, TR 6432-33; S Exh 21 at 2-3, TR 6523-24.
Generation of I'nformation' Relevant to Commercial'LMFBRs 228. The size, or the gross pcuer rating (975 MWt, 325 MWt per loop), of the CRBRP was selected as a reasonable midpoint between FFTF (400 MWt, 133 MWt per loop) and commercial size reactors (2400-3800 MWt, 600-1270 MWt per loop).      S Exh 7 at 8-13; S Exh 8 at 8-4.      Extrapolations of size by a factor of 2.5 to 3.5 are considered to be a prudent
    */
Intervenors' concern seems not to be that the timing objec-tive cannot be met but rather, in their words, that the tim-ing objective is meaningless because there is no way that the project could not meet it. I Counsel, TR 6330-31.        As mentioned, supra, Finding 191, the Commission has ruled that the validity of the timing objective itself is not open to question in this proceeding. Furthermore, the ability to meet the timing objective'is not considered in a vacuum but in concert with the other programmatic objectives. AW Longenecker, TR 6319.
 
                                      - F-172 -
compromise between the need for advancement in technology and keeping the scale-up risks acceptably low. Development of LWR technology followed approximately the same path and foreign LMFBR programs have utilized similar extrapolation 8 actors. A Exh 58 at 27, TR 6433.
229. The next plant under development by DOE and U.S. electric utilities and private industry in the LMFBR program is the Large Developmental Plant (LDP), a 1000 MWe or 2550 MWt
        " loop-type' plant. The CRBRP systems designs have already provided direct information relevant to the design of the LDP, inasmuch as the bulk of the LDP systems are based on the CRBRP systems designs.      A Exh 58 at 27-32, TR 6433-38.
230. In addition to similarities at the system level, there are strong similarities between CRBRP and LDP at the subsystem
                                */
and component level.--    The transfer of this information from the CRBRP to the LDP indicates that the CRBRP can rea-sonably be expected to provide significant information of relevance to commercial LMFBR's of the future.      A Exh 58 at 33, TR 6439.
I
    */  For example, in the reactor system the LDP reactor core is of heterogeneous design, as is the CRBRP core. The fuel material, structural material, fuel assemblies, blanket assemblies, shield assemblies, control assemblies, control rod drive mechanisms, upper internals structure, core l
restraint, instrument 3 tion, reactor head and shielding are essentially identical. A Exh 58 at 33, TR 6439.
 
                                          - F-173 -
231. A significant contribution of relevant information from the CRBRP to future LMFBR's is independent of similarities in CRBRP has already provided and will plant characteristics.
continue to provide a strong base of technological informa-tion concerning such matters as materials properties, analytical methods (e.g., thermal hydraulic analysis codes)
S A Exh 58 at 33-34, TR 6439-40; and associated data bases.
Exh 21 at 14-17, TR 6535-38.
232. The heterogeneous core configuration as used in CRBRP, including the design of the core assemblies, blanket assemblies, shield assemblies, and control assemblies and core restraint, is expected to be adopted in future LMFBR' s .
The methodology developed for heterogeneous core analysis will be directly applicable to design of larger
                        */
LMFB R' s .-"
A Exh 58 at 34-3 5, TR 6440-41; S Exh 21 at 14-15, TR 6535-36.
in CRBRP, such 233. The major engineered safety features (ESF's) as reactor containment, the liners in the cells containing
'              sodium piping, and features to mitigate the effects of l
Extensive tests of the CRBRP heterogeneous core configura-l      */
      ~~
tion were These conducted      the Zero Power Plutonium Reactor atprovided                                        ,
tests          valuable feedback on the                (
(2PPR).                            As a result of this exper-            l I
validity ience, of analytical tools.the core design of the ALDP Exhand 58 larger at    LMFBR'sj proceed with a higher degree of confidence.                                j 34-35, TR 6440-41.
 
                                - F-174 -
sodium spills and fires are all relevant to larger or commercial LMFBR's. The types of events against which these ESF's must be designed are characteristic of the LMFBR, regardless of size. Design, construction, testing, and operation of these engineering safety features will demon-strate the acceptability of these features and provide relevant information for future LMFBR's. A Exh 58 at 35, TR 6441; S Exh 21 at 13-16, TR 6534-37.
234. On the bases of Findings 228 - 233 supra, and the uncon-tradicted evidence in the record, CRBRP is likely to provide substantial information relevant to further commercial LMFBR's.
Alternative Design' Features 235. Intervenors have argued that certain alternative design fea-tures may be substantially better than those in the CRBRP.
These alternative design features are:    (1) the pool-type primary system configuration, (2) use of flywheels on sodium pumps, (3) lower system operating temperatures, (4) third shutdown system, (5) core catcher, and (6) no-vent contain-ment. A Exh 58 at 37, TR 6443.
 
i                                                              - F-175 -
i                                                      */
236. Pool-type systems -- have been considered since the very early period of LMFBR development.                The current generation of LMFBR plants includes both loop (SNR-300, BN-350, Joyo and Monju) and pool (Phenix, PFR, Superphenix, BN-600).
Recent evaluations performed in the United States have indi-cated no clear superiority of one system over the other in terms of safety, maintainability, cost and duration of fabrication and construction, and economy of operation.                On a purely functional basis, both pool and loop-type LMPBR's are feasible and neither has a significant overall advantage over the other.                There is, however, a lack of large pool-type reactor construction experience in this country, and there is a schedule risk associated with the greater esti-mated field labor require +2nts for a pool-type reactor.                AW Longenecker, TR 6363.                Therefore, there is no discernible advantage of the pool concept over the loop concept, and CRBRP in a loop plant configuration has a higher likelihood
                -*/        In a " loop-type" configuration, such as CRBRP, the major primary heat transport system components are interconnected with the reactor vessel by means of coolant-carrying piping.
In a " pool-type" configuration, the primary system compo-nents are in a " pool" of sodium contained within a vessel which also houses the reactor core. Many features, however, (e.g., intermediate heat transport system (IHTS), steam generator system (SGS), the turbine generator, and auxiliary systems), are common to both concepts. Therefore, much of the information obtained from a loop plant such as CRBRP, including contributions to the overall technology base, is relevant to either concept.            A Exh 58 at 37-38, TR 6443-44.
i
 
                                    - F-176 -
of meeting its objectives and timing.      A Exh 58 at 37-39, TR 6443-45; S Exh 7 at 8-11; S Exh 8 at 8-4.
237. Sodium pump flywheel's are not pcrt of the CRBRP,      The CRBRP primary flow coastdown characteristics (the flow vs. time after power is removed from pumps) have been selected by balancing two competing requirements:
: a. The need to' provide adequate coolant flow to the core and radial blanket for all design basis events including postulated loss of power to all three primary pumps, and;
: b. The need to minimize ths thermal transients associated with reactor and plant trips.
Too little flow might result in inadequate core cooling, while too much flow might result in overcooling and thermally stressing plant components during transients.      The required flow coastdown characteristics for the CRBRP sodium pumps are provided by building directly into the pump drive motor (as opposed to the addition of a separate flywheel) l sufficient inertia so that the required momentum of the l      pump-drive motor assembly will be available.      This inertia satisfies both of the above requirements.      A Exh 58 at 39-40, TR 6445-46; A W Anderson, TR 6364-68.
[
i l
l l
 
                                  - F-177 -
238. The addition of a heavy flywheel would be ineffective in significantly reducing the likelihood or consequences of beyond design basis events. For the postulated transient overpower (TOP) events that assume failure of both reactor shutdown systems, there would be no advantage to a heavy flywheel because the pumps continue to run in that event.
For the postulated loss of flow (LOF) events that assume failure of both reactor shutdown systems, the time for initiation of boiling would increase slightly, but once boiling is initiated, the sequence of events is controlled by the phenomena related to boiling, which are not affected by a flywheel. Increased pump inertia produced by the flywheel would not change the likelihood of sodium boiling and the resultant consequence of a non-energetic core melt-down. On the other hand, increasing the pump inertia by means of a flywheel beyond that required to provide adequate coolant flow increases the rate of temperature change asso-ciated with system thermal transients, thereby adding to the fatigue damage associated with transients.          Thus, adding a pump flywheel would not be a substantially better design alternative than the CRBRP design. A Exh 58 at 39-41, TR 6445-47.
239. The system operating temperatures of the CRBRP were selected based upon plant performance analyses that considered equip-ment constraints, steam conditions, desired fuel performance, i
 
                                - F-178 -                          l thermal transient and creep effects and cycle efficiency.
For normal operations and accidents within the design basis, lowering the operating temperatures without lowering the design temperatures would have the effect of increasing equipment sizes and costs and decreasing efficiency, while providing more margin to system limiting conditions and slightly improved fuel performance.      However, at any given design temperature, the prudent designer would provide the same structural design margins between operation and design temperatures. Thus, there is no net benefit to be derived from lower operating temperatures.      A Exh 58 at 41, TR 6447. In regard to events beyond the design basis, the effect of choosing a lower plant operating temperature would not significantly change the transient overpower hypothe-tical core disruptive accident (HCDA) consequences because the currently assumed transient overpower scenario results in molten fuel release from the pin before coolant boiling occurs. The effect of lower operating temperatures on the likelihood and consequence of a loss-of-flow HCDA is similar to that described for pump inertia selection.      The time to initiate boiling would be slightly increased, but the-like-lihood or consequences of sodium boiling would not change.
A W Anderson, TR 6313.      Lower CRBRP operating temperatures would not be a substantially better alternative for meeting project objectives.      A Exh 58 at 41-42, TR 6447-48.
i I
 
                                  - F-179 -
240. In regard to a third shutdown' system, it should be empha-sized that the CRBRP has two redundant, diverse, and inde-pendent control rod systems. S Exh 1 at II-7. A third shutdown system is unnecessary because all credible failure modes are addressed by the primary and secondary shutdown systems. A Exh 1 at 11-25, TR 2000-2024. A third shutdown system would not address any other known failure modes, and there would be no significant reduction in risk. Therefore, the addition of a third shutdown system would not be a substantially better alternative. A Exh 58 at 42-43, TR
                */
6448-49.-
l
  -*/  During the course of the hearings, Intervenors questioned Staff as to whether self-actuated shutdown systems could be available for the use in the CRBRP. While Staff Witness Long was aware of the use of such systems in the. French Phenix reactor, S W Long, TR 6469, he saw no need for them in the CRBRP in order for it to meet its programmatic objec-tives S W Long, TR 6491-92. No evidence was presented indi-cating that a Phenix-type shutdown system would be a sub-stantially better design feature.
 
                                  - F-180 -
                                      */
241. In regard to a core catcher,- although the overall approach to CRBRP design has been to include such features that make the likelihood of a core melt so unlikely that one need not include a core melt in the spectrum of design basis acci-dents, the Applicants, nevertheless, have provided margins and design features in CRBRP to mitigate the consequences of HCDA's and to assure that residual risks from,HCDA's are acceptably low. A Exh 1 at 15-26, 57-67, TR 2004-15, 2046-56; S Exh 8, Appendix J. There is no substantial additional advantage to inclusion of a core catcher in the design.        A Exh 58 at 43-44, TR 6449-50; A W Anderson, TR 6314-15, 6401-02; S W Long, TR 6492-95.
242. In regard to a no-vent containment, three situations have been considered--normal operations, design basis accidents and beyond design basis accidents.      During normal operation, the containment is continuously vented. This provides for l
  -*/  Core catcher is the name associated with the features in a plant design that would provide for the ability to retain some or all of the core subsequent to an over-power or undercooling accident that results in melting of the core and subsequent melt through of the reactor vessel and-guard vessel. A core catcher is generally assumed to include means for keeping this core debris from penetrating further into the bottom of the reactor cavity. It should be noted that (a) the core catcher does not in any way reduce the likelihood of an HCDA and that (b) any active features pro-vided in the core catcher have to perform in an extremely hostile environment subsequent to an HCDA and are inacces-sible at a time when they are required to function. A Exh 58 at 43, TR 6449.
 
                                - F-181 -
access to the containment during operation thus improving operability and maintainability of the plant.      A W Kaushal, TR 6315. In the event that any significant radioactivity levels are detected in the containment effluent, the con-tainment atmosphere is isolated through the use of contain-ment isolation valves.      Under such circumstances, the con-tainment is essentially unvented and for all design bases events may be kept unvented for as long as desired.        A Exh
                                                        */
58 at 44-45, TR 6450-51; S Exh 1 at II-13-II-18.-        In the event of a beyond-the-design-base HCDA, the containment can be vented through a clean-up system in order to maintain the containment pressure within the containment vessel capa-bility. Through the use of a clean-up system the radiological releases for such accidents can be controlled and the consequences made acceptably low.      Although design measures could be taken to increase the probability that no venting would be required, one cannot in practice foresee all contingencies nor design a perfect containment.
Therefore, as an additional margin of safety, a filtered
-*/  The CRBRP design has provision for filtering and cleanup of the vent discharge from the containment during normal opera-tion. Even with venting, the radiation dose guidelines are not exceeded. Thus, the health and safety of the public is assured even    with a vented containment. On the other hand, elimination of venting during normal operation makes the containment access during normal operation (operability, maintainability) difficult.      A Exh 58 at 45-46, TR 6451-52; S Exh 1 at II II-18.
 
                                      - F-182 -
controlled ve'nt capability was designed into the CRBRP containment to assure that containment integrity cannot be challenged. A no-vent containment is not a substantially better alternative. A Exh 58 at 46, TR 6452; S Exh 1 at II II-19.
243. Based on the uncontested evidence in the record, there are no design features which have been identified in either the U.S. LMFBR Program or in the designs utilized in foreign programs which are substantially better alternatives for satisfying CRBRP project objectives than those features incorporated in the CRBRP design. Findings 235 - 242, supra; A Exh 58 at 35, TR 6441; S Exh 21 at 29, TR 6550. ; S Exh 7 at 8-16; S Exh 8 at 8-6.
 
l l
                                            - F-183 -
Contentions'll'b)'and 11'c) (Genetic'and Somatic Effects of'CRBRP Operation) 244. Intervenors' Contentions ll(b) and (c) allege the following:
: 11. The health and safety consequences to the public and plant employees which may occur if the CRBR merely complies with current NRC standards for radiation protection of the public health and safety have not been adequately analyzed by Appli-cants or Staff.
(b)  Neither Applicants nor Staff have adequately assessed the genetic effects from radiation exposure including genetic effects to the general population from plant employee exposure.
(c)  Neither Applicants nor Staff have adequately assessed the induction of cancer from the exposure of plant employees and the public.
Board Order of April 14, 1982, Appendix I at 14-15.      Inter-venors' Contentions ll(b) and (c) question whether the Staff or Applicants have adequately analyzed the genetic and somatic risks resulting from normal operation of CRBRP.
245. In calculating both the genetic and somatic effects, Appli-cants assumed an occupational exposure of 400 man-rem,      a l
general population exposure of 0.1 man-rem, and a population of 921,200 persons within a 50 mile radius of CRBRP.        A Exh 42 at 4, TR 4270; S Exh 8 at 5-20; A Exh 34, Vol.      1, Section 2.1.5; A Exh 35, Vol. 2,  Chapter 5.2 at 5.2-19-5.2-41.
l l
246. In considering genetic effects, Applicants used the back-ground frequencies for the various genetic effects in the BEIR III Report. This report was also used to estimate the increase in the frequencies of the different genetic end-
 
l
                                            - F-184 -
points as a result of radiation exposure.      A Exh 42 at 5, TR 4271. The BEIR III estimates are the most appropriate basis for estimating the genetic ef fects from operation of CRBRP.
A Exh 42 at 5, TR 4271; S Exh 12 at 6, TR 4117.        The BEIR III estimates are conservative and likely to overestimate the genetic effects.      S Exh 12 at 7, TR 4118; S W Bender, TR 4069-4071. In addition to the conservatism inherent in the BEIR III estimates, Applicants and Staff made a number of conservative assumptions, thus ensuring that the upper esti-mates calculated for the various genetic disorders are
                                  */
maximal values.--      A Exh 42 at 25, TR 4291; S Exh 12 at 7-8, TR 4118-4119.
247. In applying the BEIR III estimates, Applicants considered all relevant classes of genetic disorders -- autosomal dominants and x-linked disorders, recessive disorders, chromosome alterations and irregularly inherited diseases, A i
l Exh 42 at 9-24, TR 4275-4290, and separately calculated the l
l                effects of CRBRP operation as to each of these disorders.        A Exh 42 at 24, TR 4290.      Applicants expressed the estimated genetic effects in terms of a range.      A Exh 42 at 24, TR 4290.
            */  For example, the whole body dose was used as an estimate of
            -~
the gonadal dose although the transuranic radionuclides such as Pu 239 do not concentrate in the reproductive organs.      A Exh 42 at 25, TR 4291; S Exh 12 at 5, TR 4116.
l
 
                                        - F-185 -
248. Based on the BEIR III estimates and the conservative assump-tions used by Applicants, the total genetic disorders in the general population per million liveborn range from 0.29 x 10-2 as the upper limit to 0.06 x 10-3 as the lower limit.
A Exh 42 at 24, TR 4290.      In contrast, the current incidence of genetic disorders per million liveborn is 106,000.      S Exh 12 at 10, TR 4121.
249. For occupational exposures, based on the BEIR III estimates and the conservative assumptions used by Applicants, the total increase in genetic disorders per 1,000 liveborn range from 1.3 as an upper limit to 0.19 as the lower limit. In contrast, the current incidence of genetic disorders per thousand liveborn is 106. A Exh 42 at 24, TR 4290.
250. In considering the somatic effects of operation of CRBRP, Applicants used the conservative, linear no-threshold hypo-thesis contained in the BEIR III Report in order to extend existing data on the number of cancers that may result from a given dose of radiation to the lower doses associated with CRBRP . A Exh 42 at 26, TR 4292. The linear no-threshold model tends to overestimate the risk. A W McClellan, TR 4022; S Exh 13 at 6, TR 4149. In addition, Applicants used
                                                */
both the absolute risk approachF- and the relative risk
        -*/  The absolute risk approach expresses the results in increased numbers of cancer cases per million person-rem.      A Exh 42 at 27, TR 4293.
l l
 
i
                                  - F-186 -
approachr# in estimating the total range of somatic i      effects. A Exh 42 at 26-27, TR 4292-4293.
251. Based on the linear no-threshold model, and using both the absolute and relative risk approach, the estimated effects range from 0.000015 to 0.00005 cancers among the public per reactor year. A Exh 42 at 28, TR 4294; A W McClellan, TR 4003. The estimated effects among the workers range from 0.07 to 0.2 cancers per reactor year.      In contrast, approxi-mately 16 percent, or one in six of all persons in the population at large, would be expected to die of cancer.      A Exh 42 at 28, TR 4294. The lower range of each of these two estimated ranges could in fact be zero.      A W McClellan, TR 4033. The upper range, in light of the conservative nature of the assumptions, is an upper bound estimate and it is very unlikely that it would be exceeded.      A W McClellan TR 4033.
252. The NRC Staff experts, although using different assumptions, calculated health effects similar in magnitude to those calculated by Applicants.      Staff's genetic expert, Dr.
Bender, who served on the BEIR III Committee, calculated combined occupational and general public genetic effects of 1.8 to 33 genetic disorders per million liveborn over all
  -*/  The relative risk approach expresses the results as a per-centage increase in normal cancer incidence per million person-rem. A Exh 42 at 27, TR 4293.
 
                                - F-187 -
time. Because 106,000 genetic disorders occur in each          ,
generation spontaneously, the first generation increase in risk caused by operation of CRBhP amounts at most to 0.00002 percent. In subsequent generations, the risk would be even less. S Exh 12 at  10,  TR 4121.-/ Dr. Bender concluded that the increase per generation would very likely be smaller than the upper limit estimate, that the increase would not be detectable, and that the genetic effects from operation of CRBRP will be so small as to constitute a insignificant impact upon human health and welfare.      S Exh 12 at 13, TR 4124; See S W Bender, TR 4106-4107.
253. In estimating somatic effects, the Staff conservatively used the BEIR I linear non-threshold dose response model.      S Exh 13 at 6, TR 4149. The risk estimators used by the Staff were consistent with those used in BEIR I, BEIR III and UNSCEAR. S Exh 13 at 6-7, TR 4149-50; S Exh 8 at 5-15.
254. The Staff calculated the risk of potential premature cancer to the maximally exposed individuals- / as 6.7 x 10-7      The risk of cancer to an individual within 50 miles of CRBRP is
*/    In the FES Supplement, the Staff calculated 9 genetic effects from both occupational and non-occupational exposure assuming 30 years operation of CRBRP, which falls within the range of 1.8 to 33 calculated by Dr. Bender. S Exh 12 at 13, TR 4124; S Exh 8 at 5-21.
  **/ The maximally exposed individual is a hypothetical      individu-
-~-
ally potentially subject to maximum exposure. S Exh 13 at 3, TR 4146.
 
l
                                              - F-188 -
much less.                S Exh 13 at 7,  TR 4150; S Exh 8 at 5-21. The risk to the public is much less than the risk from exposure to other sources of radiation such as medical exposure or natural background radiation.                  S Exh 13 at 10, TR 4153. As for the occupational work force, the Staff calculated, assuming 1,000 exposed individuals, that 0.14 additional cancers per reactor year would result S Exh 13 at 9, TR 4152. The current incidence of cancer per 1,000 workers is 160. S Exh 13 at 8, TR 4151; S Exh 8 at 5-15.
255. Intervenors failed to introduce any evidence regarding Contentions ll(b) and (c).                  Intervenors' major concerns regarding Staff and Applicants' analyses apparently are (1) that uncertainties in the BEIR III Report were i.ot taken into account, (2) that the views of other experts were not taken into account, and (3) the recent studies of the neutron / gamma dose contributions at Nagasaki and Hiroshima were not considered.
256. As to the uncertainties in the BEIR III Report, the uncon-tradicted record evidence demonstrates that the values calculated by Applicants and Staff in considering genetic and somatic health effects were upper bound limits.                AW McClellan, TR 4033; A Exh 42 at 25, TR 4291; A Exh 42 at 27-28, TR 4293-4294; S W Bender, TR 4071-4084.                Both Applicants and Staff made conservative assumptions which had the effect of overestimating the expected health effects from operation
 
                                                            - F-189 -
                    */
o f CRB RP .--      A Exh 42 at 25, TR 4291; A Exh 42 at 26, TR 4292.
257. Both Applicants and Staff relied upon well-qualified experts in the fields of genetic and somatic health effects due to exposure to radiation.                              A Exh 42 at 30-37, TR 4296-4303; S Exh 12, TR 4125.                    In performing their analyses, these experts used data and methodologies which are widely accepted in the scientific community.                              A Exh 42 at 5,  26-27, TR 4271, 4291-4292; S Exh 12 at 2, 6-7, TR 4113, 4117-4118; S Exh 13 at 4-7, TR 4147-4150; S W Bender, TR 4068-4071; A W McClellan, TR 4021-4022. In addition, these experts considered the
                                                                                    **/
contrary views of Gofman, among other experts .---                                Dr.
Bender, an expert in genetics, noted that he did not consider Gofman an expert in genetics and that Gofman misunderstood various genetic principles.                              S W Bender, TR 4095. In addi-tion, Applicants' testimony addressed Dr. Gofman's views and showed that several of his hypotheses concerning genetic effects had no merit.                              See A Exh 42 at 11, TR 4277.
258. The recent reevaluations of the neutron and gamma doses at Nagasaki and Hiroshima do not create any substantial uncer-tainty with regard to the analysis of health effects.- The
    */    The only uncertainty revealed by Intervenors' cross examina-
  ~~
tion was in regard to the linear quadratic model in BEIR III. Neither Staff nor Applicants used this model, however.
A W McClellan, TR 4017-4025.
    **/  See also Findings 58-60 supra.
 
1
                                                          - F-190 -
uncontradicted evidence shows that no substantial changes in the BEIR III Report risk estimators are expected from the reevaluation ,-                            A W Thompson, TR 4029; S W Bender, TR 4075-76, S Exh 12 at 7-8, TR 4118-19.
259. Applicants and the Staff have adequately analyzed the genetic and somatic effects of CRBRP operation, and have demonstrated that the risk of increased health effects due to operation of CRBRP is not significant.                            See Findings 245-258.
    -*/  For the most part, Intervenors' questions regarding the re-evaluation of the Nagasaki and Hiroshima doses were directed to the linear quadratic model in BEIR III. As noted earlier, this model was not used by either Applicants or Staff. See TR 4029-4030.
 
                                                                                                        - F-191 -
ULTIMATE ENVIRONMENTAL FINDINGS 260. The Final Environmental Statement (FES) (S Exh 7) and Final Supplement to the Final Environmental Statement (FSFES) (S Exh 8) have been introduced into evidence and fully consid-ered by the Board.
261. The environmental effects of CRBRP accidents, including Class 9 accidents, have been adequately analyzed, are extremely low, and are comparable to those of LWR's.                                    Find-ings 72-87, sopra; See Findings 10-46, supra.
262. The risks to national security and national energy supply attributable to CRBRP accidents have been adequately analyzed and are negligible.                                  Findings 88-104, supra.
263. The environmental costs (radiological risks and economic costs) of safeguarding CRBRP and its supporting fuel cycle have been adequately analyzed and are extremely low.                                The economic costs are small fractions of the total plant costs, and would not significantly affect the cost-benefit balance for CRBRP.                            Findings 105-159, supra.
264. The environmental impacts of CRBRP fuel cycle activities have been conservatively analyzed and constitute a substan-tial overestimation of the expected environmental impacts.
Findings 160-188, supra.
265. All reasonable alternative sites and siting concepts have been adequately analyzed and none are substantially better
 
J
                                            - F-192 -
alternatives to the CRBRP as proposed.                                Findings 190-209, supra.
266. The CRBRP is likely to meet its programmatic objectives and to generate information of substantial value and relevance to the LMFBR program and future LMFBR's.                                                All reasonable alternative design concepts have been adequately analyzed and none are substantially better than the CRBRP as pro-posed. Findings 210-243, supra.
267. The genetic and somatic effects of CRBRP operation on the public and plant workers have been conservatively analyzed, and constitute a substantial overestimation of the expected effects.          The risks of increased health offects due to operation of CRBRP is not significant.                                  Findings 244-259, supra.
268. The Board has decided all matters in controversy among the parties within the scope of NEPA.                    Findings 72-259, supra.
269. The FES and FSFES utilized a systematic interdisciplinary approach and integrated the pertinent natural, physical, social, and environmental design arts. S Exh 7; S Exh 8, Appendix M.
270. The FES and FSFES are detailed statements which adequately analyze:              a) the environmental impacts of the proposed CRBRP (S Exh 7, Sections 4, 5, 7, and 10.1, Appendices D and E; S Exh 8 Sections 4,            5,  7, 10.1, Appendices D, E, H, and J; Findings 72-259, supra); b) any adverse impacts which cannot
 
                                          - F-193 -
be avoided should CRBRP be constructed and operated (S Exh 7, Section 10.1; S Exh 8, Section 10.1; Findings 72-259, supra); c) alternatives to CRBRP (S Exh 7, Section 9; S Exh 8, Section 9 and Appendix L; Findings 190-243, supra);
d) the relationship between local short-term uses of the l          environment and the maintenance and enhancement of long-term productivity (S Exh 7, Section 10.2; S Exh 8, Section 10.2);
and e) any irretrievable and irreversible commitments of resources which would be involved in construction and opera-tion of CRBRP . (S Exh 7, Section 10.3; S Exh 8, Section 10.3.).
271. Construction and operation of CRBRP will yield substantial informational benefits for the LMFBR program. S Exh 7, Section 10.4.1; S Exh 8, Section 10.4.1; Findings 210-243, supra.
272. The FES and FSFES were circulated for comment and the NRC Staff considered, incorporated, and properly resolved comments from Federal, State and local agencies which have special expertise, interest, and regulatory jurisdiction with respect to CRB RP , and from interested members of the public. S Exh 7, Appendix A; S Exh 8, Sections 11, 12, and Appendix A. The FES and FSFES were made available to the President's Council on Environmental Quality, and were considered at all pertinent stages of the NRC review
 
                                                    - F-194 -
process, including the CRBRP LWA hearings.            S Exh 7 at lii; S Exh 8 at v.
273. Upon consideration of all of the factors contained in the record, including all matters in controversy between the parties, the Board finds that the FES and FSFES properly analyzed the environmental, economic, technical and other benefits of CRBRP and the environmental and other costs of CRBRP, and upon consideration of these benefits and costs, further finds that the, benefits substi.ntially outweigh the costs.      S Exh 7, Section 10.4.3; S Exh 8, Section 10.4.3.
274. The Findings herein affirm the analyses and conclusions of the FES and FSFES.        Findings 72-273, supra.
l l
i - _ _ .  . _ _ _ . _ _ _ _ . . __ --_-_-_        __-  . _ .  .        . _ _ _ _
 
C'- 1 VII. CONCLUSIONS OF LAW The Board has considered all of the evidence submitted by the pc.rties and the entire record of this proceeding. Based upon the Opinion and Findings of Fact set forth herein, which are supported by reliable, probative and substantial evidence in the record, this Board, having decided all matters in controversy, concludes:
: 1. The Clinch River site is suitable for a reactor of the general size and type proposed in the application from the standpoint of radiological health and safety considerations.
: 2. The LWA procedure set forth in 10 C.F.R. { 50.10(e) is applicable to CRBRP.
: 3. The content of the Final Environmental Statement and Final Supplement to the Final Environmental Statement (S Exh's 7 & 8) is hereby affirmed.
: 4. The requirements of NEPA, 42 U.S.C. I 4321 et_ seq., and 10 C.F.R. Part 51 have been complied with in this proceeding.
: 5. Upon balancing all conflicting factors contained in the record of the proceeding, and weighing the environ-l
 
C-2 mental, economic, technical and other benefits against the environmental and other costs, and considering available alternatives, the Board concludes that a limited work authorization should be issued for the CRBRP pursuant to 10 C.F.R. $ 50.10(e).
l
                                                        =
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l i
APPENDIX A EXHIBIT LIST l
l
              ?
f
)
 
l EXHIBIT LIST I. Applicants' Exhibits Marked l
Exhibit                                        for I.D. Offered  Admitted No.      Description 1      Applicants' Testimony Concerning      1282    1986      1988 NRDC Contentions 1,2, and 3.
1282    2097      2116 2-PSAR Section 2.3, Meteorology.
3 PSAR Section 4.2.3, Reactivity      1282    2097      2116 Control Systems.
4 PSAR Section 5, Heat Transport and    1282    2097      2116 Connected Systems.
5 PSAR Section 6.2, Containment        1282    2097      2116 Systems.
6 PSAR Section 7.1.2, Identification    1282    2097      2116 of Safety Criteria.
1282    2097      2116 7
PSAR Section 7.5.4, Fuel Failure Monitoring System. ,
8      PSAR Section 15.1.1, Design          1282    2097      2116 Approach to Safety.
9 PSAR Section 15.1.4, Effect of Design Changes on Analyses of        1282      2097      2116 Accident Events.
10      PSAR Section 15.2, Reactivity          1282    2097      2116 Insertion Design Events.
11      PSAR Section 15.3, Undercooling      1282    2097      2116 Design Events.
12      PSAR Section 15.4, Local Failure      1282      2097      2116 Events.
 
Marked                  Admitted Exhibit                                              for I.D. Offered No. _  Description 2097        2116 1282 13 PSAR Section 15.6, Sodium Spills.
14 PSAR Section 15A, Radiological Source Term for Assessment of                      2097        2116 1282 Site Suitability.
l 15 CRBRP-3, Vol.1, Section 4.0-4.4 and App. 4A, Assessment of HCDA                    2097        2116 1282 Energetics.
16 CRBRP-3, Vol. 1, Sections 5.0-5.5, Assessment of Structural          1282      2097        2116 Margin Beyond the Design Base.
17 CRBRP-3, Vol.2, Sections 2.0-2.4, Design Features Providing Thermal                  2097        2116 1282 Margin Beyond the Design Base.
18 CRBRP-3, Vol.2, Sections 3.0-?.4,      1282      2097        2116 Assessment of Thermal Margin.
19 CRBRP-3, App.A, Development Programs Supporting Thermal                        2097        2116 1283 Margin Assessments.
WARD-D-0185, Primary Piping                        2097        2116 24                                              1283 integrity Report.
Applicants' Testimony Concerning                  2072        2073 25                                              1284 NRDC Contention 2(e).                                          2073 1892      2072 28      Errata Sheet to Exhibit 25.
29 The Consequences of Safety Prescriptions for Fast Breeder        2646      2794        2798 Reactor Design in France.
incidents and Accidents Considered                2794_      2801 30                                              2694 i
'                                in the Safety Analysis of CDFR.
31      Design Criteria, Concepts and Features Important to Safety in        2696      2794        2801 Licensing.
l
 
Marked Exhibit                                        for I.D.      Offered Admitted i            No. Description 33    Letter from Leslie Silverman to John A. McCone, dated Dec.10,          2985        3049            3148 1960.
3240        3240              3241 34-38 Environmental Report (5 volumes).
39    Applicants' Testimony Concerning Safeguards (NRDC Contentions 4            3246        3472              3473 and 6. b. 4) .
40    GAO Report " Obstacles to U.S.
l Ability to Control and Track Weapons-Grade Uranium Supplied            3810        3864              3864 Abroad," GAO-ID-82-21, Aug.2, 1982.
42    Applicants' Testimony Concerning          3995        4266                4266 NRDC Contentions ll(b) and ll(c).
43    Applicants' Testimony Concernin    .      4142        4323                4323 NRDC Contentions 6.b.1 and 6.b.
45    Applicants' Direct Testimony Concerning Intervenors' Contentions        4625        4731                4732 Sa) and 7c).
46    App. Testimony Concerning Conten-tions 2d), 2f), 2g), 2h), 3d)
(Environmental Effects) and 5b)            4965        5345                  5374 (with glossary) .
47    App. Testimony Concerning Conten-          4965        5345                  5374 tion 5b) (with glossary).
48    Excerpts from Rocky Flats FES              5838        6013                6016 (April 1980).
49
                        " Chromosome Changes'in So=atic Cells of Workers with Internal Dispositions of Plutonium;"
5855        6013                  6016 Brandom, et al. IAEA-SM-237/38 50    " Dose-Rate Conversion Factors for External Exposure to Photon and Electron Radiation from Radio-nuclides Occuring in Routine Releases from Nuclear Cycle Facilitiess" Kocher, Vol. 38              5863          6013              6016 Health Physics, 1980
 
i 4-Marked Exhibit Description                                      for I.D. Offered Admitted No.
51    Copy of Transparency, " Histograms of the Prevalence of Structural                                            f Chromosome Aberration in Rocky Flats Controls and Plutonium                    5871    6013    6016 Workers."
52    Letter from William A. Mills, Acting Dep. Asst., Administrator for Radiation Programs, EPA to                            6013  6016 Carl J. Johnson (2/27/79).                      5886 53 -  NUREG/CR-1659, " Reactor Safety Study Methodology Applications Program; Calvert Cliffs No.2, PWR Power Plant,"                            6289 6118    6289 (May 1982).
54    " Primary Containment Leakage Integrity Availability"and            Review of Failure Weinstein, Nuclear Experience, Safety, Vol. 21, No. 5 (Sept./Oct.              6140    6289    6289 1980).
55  Extract from Appendix VI of WASH-                        6289    6289 1400.                                            6150 56  Letter from B. Johnson, P. Davis                        6289    6289 and H. Lee to Morris Udall (2/21/79). 6162 57  Excerpts from " Final Report on Comparative Calculations for the AEC and CRAC Risk Assessment Codes,"                              6289 6166    6289 SAI (Dec. 1978).
58    Applicants' Testimony Concerning                6293    6405    6406 Contentions 7a) and 7b).
u  -    - , - - - - - - - - - -
                                            -yyw
 
II. Staff Exhibits Marked Exhibit                Description                                for I.D.      Offered        Admitted No.
i 1740      2444          2444 1                    CRBRP Site Suitability Report.
2                  Staff Testimony on Intervenors' Contentions 1(a), 2(b), 3(b),                                2444        2444 3(c), and 3(d).                                  2122 3                    Staff Testimony on Intervenors' Contention 2(a), 2(c), 2(d), 2(e),                            2444        2444 2(f), 2(g) and 2(h).                            2123 4                  ACRS Report on CRBRP Site                                    2444        2444 Suitability    dated July 13, 1982.              2123 5                  Letter from Richard Denise to                    2277        2444        2444 Lochlin Caffey dated May 6, 1976.
6                  Staff's Standard Review Plan,                                3191        3192 3192 Section 2.2.3.
3243        3243      3244 7                  1977 CRBRP FES (NUREG-0139).
8                  Supplement to CRBRP FES (October                              3243        3244 3243 1982).
3243        3243      3244 9                  Errata Sheet to FES Supplement (Nov. 15, 1982).
10                    Staff Testimony Regarding Contentions                          3731      3732 3565 4 and 6(b)(4).
11                    Staff Testimony Regarding Contentions                          3704. 3704 3701 c
4 and 6(b)(4).
12                    Testimony of Michael A. Bender                    4063          4111    4111 I
Regarding Contention ll(b).
13                  Testimony of Edward F. Branagan, Jr. 4063                        4142      4142 Regarding Contention ll(c).
l 14                  Staff Testimony Regarding Contention                            4443    4443 4349 6.
 
Marked Exhibit Description                        for I.D. Offered Admitted No.
15      Staff Testimony on Contentions 5(a)          4864    4864 4756 and 7(c).
16      Testimony of Paul Leech on Contention          4864  4864 4761 7(c).
17      Staff Testimony on Contentions 2c,            5650  5747 2d, 2f, 2g, 2h, 3c and 3d.          5442 5653    5682  5682 18      Staff Testimony on Contention 5b.
5324    5324  5324 19      Errata Corrections to NUREG-0139.
20      "A Note on the Pipe Rupture Probability Calculations for the Primary Heat Transport System of    6172    6289    6289
'          CRBRP," Harris, SAI (10/7/77).
21      Staff Testimony Relative to          6463      6521  6521
'          Contentions 7(a) and 7(b).
e i
l J
l
 
III. Intervenors' Exhibits Marked Exhibit                                          for I.D. Offered  Admitted No. Description I
3    Testimony of Dr. Thomas B. Cochran,        ---      2809      2809 Part I.
4    Testimony of Thomas B. Cochran,            ---      3050      3050 Part II.
5      Safety Measures at the Creys-            2733      2801      2802 Malville Power Station.
6    Super Phenix News, July, 1978,            2733        2801      2802 No. 1.
2733        2801      2802 7    Super Phenix News, March, 1982, No. 7.
2875        3050      3050 8    Testimony of Dr. John C. Cobb.
2879        3050      3050 9    Testimony of Dr. Karl Z. Morgan.
10A    Extracts from Proposed Guidelines on Dose Limits for Persons Exposed to Transuranium Elements in the            3189      3189      3189 General Environment.
11    GAO Report-Nuclear Fuel Reprocess-ing and the Problems of Safeguard-ing Against the Spread of Nuclear          3316      3473      3562 Weapons.
12    Testimony of Dr. Thomas B. Cochran, Part V (Intervenors' Contentions 4        3756      3886      3886 and 6(b)(4)).
12A    Supplement to Testimony of Dr.            3760      3886      3886 Thomas B. Cochran, Part V.
13    Testimony of Dr. Thomas B. Cochran, Part III (NRDC Contentions 6(b)(1)          4476      4566    4566 and (3)).
NUREG 0002, Vol. 3, GESMO Report 14                                                4617      4617      4617 pp. 4c-24.
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III. Intervenors' Exhibits Exhibit                                            Marked Description                              for I.D. Offered Admitted No.
5810  6003      6017 21    Testimony of Carl J. Johnson.
22    Testimony of Dr. Cochran, Part IV            6093    6193    6194 (Accidents).
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APPENDIX B WITNESS LIST
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WITNESS LIST
: 1. Applicants' Witnesses Transcript Pages Namq Position 1293-1295.1352-Licensing Specialist, Westinghouse  1393,1461-1881, Nail W. Brown          Corp.                                1989-2071 1292-1295, 1352-Manager of Licensing for the CRBRP    1393,1461-1881, George H. Clare        Project, Westinghouse Advanced        1989-2071 Research Division.
1292-1295, 1352-Associate Director, Reactor          1393, 1461-1881, Livrence W. Deitrich  Analysis and Safety Division,        1989-2071 Argonne National Laboratory.
1293-1295, 1352-Technical Asst. to the CRBRP          1393, 1461-1881, Vcncil O' Block        Systems Integration Manager,          1989-2071 Westinghouse Advanced Reactor Div.
1292-1295, 1352-Lee Strawbridge Manager of Nuclear Safety and        1393, 1461-1881, Licensing, Westinghouse Advanced      1989-2071 Reactor Div.
Staff Member, Los Alamos National    1884-1922, 2074-John W. Healy          Laboratory.                          2091 1884-1922, 2074-President and Director, Inhalation    2091 Roger O. McClellan    Toxicology Research Institute, Lovelace Biomedical and Environ-mental Research Institute.
1884-1922, 2074-Senior Staff Scientist, Pacific      2091 Roy C. Thompson        Northwest Laboratory.
                                                                  =
Office of Safeguards and Security,    3246-3560 Glenn A. Hammond      DOE.
3247-3560 Industrial Security Specialist.
Edward F. Penico President and Director, Inhalation    3994-4057, Roger O. McClellan    Toxicology Research Institute.        4267-4303
 
V-i Transcript Pages Nrma                    Position Staff Member, Los Alamos National    3994-4057, John W. Healy                                                  4267-4303 Lab.
Roy C. Thompson Senior Staff Scientist, Pacific      3994-4057, Northwest Lab.                        4267-4303 R. Julian Preston        Senior Research Staff Member, Oak      3994-4057, Ridge National Lab.                    4267-4303 4158-4266, Gzorge L. Sherwood,Jr. Nuclear Engineer, DOE.                  4304-4347 4158-4266, Douglas C. Newton        Nuclear Engineer, DOE.
4304-4347 William M. Hartman      Manager, LMFBR Fuels Supply and      4158-4266, Process Development, DOE.
4304-4347 Orlan O. Yarbro          Program Manager, Oak Ridge National    4158-4266, Lab.                                  4304-4347 Lawrence J. Kripps      Senior Analyst, Energy Incorporated.4226-4752 Gzorge H. Clare          Manager of Licensing, Westing-        4966-5344, house CRBRP.                          5375-5420 Lee F. Strawbridge        Manager, Nuclear Safety and          4966-5344; Licensing, Westinghouse.              5375-5420 Lawrence W. Deitrich    Associate Division Director,          4966-5344; Argonne National Laboratory.          5375-5420 i
H. Wayne Hibbitts        Chief, Safety and Environmental          -
4966-5344; Branch, CRBRP Project Office.
5421-38 John R. Longenecker      Acting Director, Office of CRBRP, Office of Breeder Reacter Programs, 6291-6460 DOE.
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!          Carl A. Anderson, Jr. Project Mgr.,Large Plant Projects,    6291-6460 Westinghouse.
Dep. Asst. Dir, for Emergency,        6291-6460 Narinder N. Kaushal CRBRP.
 
/
II. Staff Witnesses Transcript Pages N-m1                    Position Nuclear Engineer, Accident        2118-2439, 2483-Lsrry W. Bell                                              2585 Evaluation Br. NRC Office of Nuclear React,r Regulation.
2118-2439, 2483-Edward F. Branagan,Jr. Radiological h.jsicist Radiolog- 2585 ical Assessment Br., NRC Office of Nuclear Reactor Regulation.
Chief of Accident Evaluation Br.,2118-2439, 2483-Lewis G. Hulcan          NRC Office of Nuclear Reactor      2585 Regulation.
Reactor Physicist, Reactor        2118-2439, 2483-John K. Long Systems Br., Nuclear Regulatory 2585 Commission.
Reactor Engineer, CRBRP Program  2118-2439, 2445-Jerry J. Swift Office, NRC Office of Nuclear      2482, 2483-2585 Reactor Regulation.
Senior Containment Systems        2118-2439, 2483-Farouk Eltawila          Engineer, Containment Systems      2585 Br., NRC Office of Nuclear Reactor Regulation.
Chief of Meteorology Section,      2118-2439, 2483-Irwin Spickler                                            2585 Accident Evaluation Br., NRC Office of Nuclear Reactor Regulation.
Technical Review Section Leader. 2118-2439, 2445-Bill M. Morris          CRBRP Program Office, NRC Office 2482 of Nuclear. Reactor Regulation.
Reactor Engineer, CRBRP Program 2118-2439, 2445-Thomas L. King                                              2482_
Office, NRC Office of Nuclear Reactor Regulation.
2118-2439, 2445-Edward T. Rumble III    Corporate Vice-President,          2482, Science Applications, Inc.
1
 
II Staff Witnesses Transcript Pages Name Position Section Chief, NRC, Division of'            3563-3646, Robert J. Dube                  Safeguards.                                  3663-3753 MCEA Program Analyst, NRC Division          3563-3646, Robert D. Hurt                  of Safeguards.                              3663-3753 Senior Safeguards Scientist, NRC,            3563-3646, John W. Hockert                  Div. of Safeguards.                          3663-3753 Plant Protection Analyst, NRC,              3563-3646, Charles E. Gaskin                Div. of Safeguards.                        3663-3753 Harvey B. Jones                    Safeguards. Analyst, NRC, Div. of          3563-3646, Safeguards.                                3663-3753 Senior Scientist, Brookhaven                4060-4127 Michael A. Bender                  National Lab.
Edward F. Branagan, Jr.
Radiological Physicist, NRC Office of Nuclear Reactor                  4063, 4134-4141, Regulation.          (Contention 11 C)    4144-4157 (Contention 6)        4348-4472 Homer Lowenberg                    Chief Engineer, Office of Nuclear Material Safety and Safeguards.            4348-4472 (Contention 6)
(Contention 5a and 7c) 4754-4906 Section Leader, NRC, Division of Regis R. Boyle                      Waste Mgt.
4348-4472 A. Thomas Clark,Jr.                  Senior Chemical Engineer, NRC, Division of Fuel Cycle and Material 4348-4472 Safety.
Site Analyst, NRC, Division of Charles Ferrell                      Engineering.
4754-4906 l
)    ---        - - - - -  . . _ _ _ _ _ _ _ _ _ .              _
 
II Staff Witnesses (Continued)
Transcript Pages Name                      Position Section Leader, NRC, Office Lotnard Soffer            of Nuclear Reactor Regulation. 4754-4906 Irwin Spiekler Section Leader, NRC Office of Nuclear Reactor Regulation.      4754-4906 Pcul Leech              Project Manager, NRC Office of    4754-4864, Nuclear Reactor Regulation.
4907-4923 Bill M. Morris            Technical Review Section Leader, Office of Nuclear        5439-5650; 5748-5800 Reactor Regulation (NRR).
Jerry J. Swift          Reactor Engineer, Office of      5439-5650; 5748-5800 NRR.
John K. Long            Reactor Engineer, Office of      5439-5650; 5748-5800 NRR.                              (Accidents); 6462-6551 (Design Alt.)
Vice President, Science          5439-5650; 5748-580C Edmund T. Rumble III Applications, Inc.
Lewis G. Hulman        Chief, Acsident Evaluation      5439-5650; 5748-580C Br., Office of NRR.
Mohan C. Thadani      Nuclear Engineer, Office of      5439-5650; 5748-580C NRR.
(Cont. 2&3) 5652-5703 (Cont. 5b)
Homer Lowenberg      Chief Engineer, Office of Nuclear Material Safety and      5652-5703 (Cont. 5b ',
Safeguards.                      6075-6086 (Rebuttal of Johnson);
Leonard Soffer        Site Analysis Section Leader,    5652-5/03 l                                  Office of NRR.
Paul H. Leech          Senior CRBRP Project Manager, 6462-6551 l
Office of NRR.
Richard A. Becker      Reactor Engineer, Office of      6462-6551 NRR.
 
III. Intervenors' Witnesses Transcript Pages Nams                  Position Thomas B. Cochran Senior Staff Scientist, Natural 2593-2792, 2810-3018, Resources Defense Council, Inc. 3051-3099 Professor of Community Health,  2874-2925, 3100-3118 John C. Cobb          University of Colorado School of Medicine.
2874-3018, 3119-3143, Kcrl Z. Morgan        Consultant on Radiation Protec-  3150-3188 tion.
Thomas B. Cochran      Senior Staff Scientist NRDC. 3755-3992 (Safeguards)
(Fuel Cycle)    4473-4616 Carl J. Johnson        Assoc. Clinical Prof., U. of      5809-6003; 6018-6074 Colorado School of Medicine.
Senior Staff Scientist, NRDC. 6086-6289 (Accidents)
Thomas B. Cochran i
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3-Lawson McGhee Public Library 500 West Church Street Knoxville, Tennessee    37902 William E. Lantrip, Esq.
Attorney for the City of Oak-Ridge Municipal Building Post Office Box 1 Oak Ridge, Tennessee    37830
                      ** Leon Silverstrom, Esq.
Warren E. Bergholz, Jr., Esq.
U. S. Department of Energy 1000 Independence Ave., S. W.
Room 6-B-256, Forrestal Building Washington, D. C. 20585 (2 copies)
Eldon V. C. Greenberg Galloway & Greenberg 1725 Eye St., N. W., Suite  601 Washington, D. C. 20006 Ccamissioner James Cotham Tennessee Department of Economic and Community Development Andrew Jackson Building, Suite 1007 Nashville, Tennessee    37219 A,                _
Georgejp[Edg'ar 'g Attorney ror      V Project Management Corporation l
DATED: January 24, 1983 l
                                                              =
L l
l  */      Denotes hand deliiery to 1717 "H'.' Street, N.W., Washington, D.
l
    **/ Denotes hand delivery to indicated address.
l
    ***/ Denotes hand delivery to 4350 East-West Highway, Bethesda, Md.
  ****/ Denotes hand delivery to 7735 Old Georgetown Road (Maryland National Bank Building,) Bethesda, Maryland.
  *****/ Denotes delivery by Air Express
 
4 I
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* Atomic Safety & Licensing Appeal Board l
U. S. Nuclear. Regulatory Commission Washington, D. C. 20545
* Atomic Safety & Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. C. 20545
* Docketing & Service Section Office of the Secretary U. S. Nuclear Regulatory Commission Washington, D. C. 20545 (3 copies)
William M. Leech, Jr., Attorney General William B. Hubbard, Chief Deputy Attorney General Lee Breckenridge, Assistant Attorney General State of Tennessee Office of the Attorney General 450 James Robertson Parkway Nashville, Tennessee  37219 Oak Ridge Public Library Civic Center Oak Ridge, Tennessee  37820 l
Herbert S. Sanger, Jr., Esquire                -
Lewis E. Wallace, Esquire W. Walter LaRoche, Esquire James F. Burger, Esquire Edward J. Vigluicci, Esquire Office of the General Counsel
  -    Tennessee Valley Authority l
400 Commerce Avenue            (2 copies) l
'      Knoxville, Tennessee  37902
    **Dr. Thomas Cochran Barbara A. Finamore, Esquire Natural Resources Defense  Council W., Suite 600 1725 Eye Street, N. 20006  (2 copies) l
'      Washington, D. C.                          _
Ellyn R. Weiss, Esquire Harmon & Weiss 1725 Eye Street, N. W., Suite 506 Washington, D. C. 20006 l
 
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
                                                                                  )
                                                                                  )
In the Matter of                                                              )
                                                                                  )
UNITED STATES DE?ARTMENT OF ENERGY                                            )
                                                                                  )  Docket No. 50-537 PROJECT MANAGEMENT CORPORATION                                                )
                                                                                  )
TENNESSEE VALLEY AUTHORITY                                                      )
                                                                                    )
(Clinch River Breeder Reactor Plant)                                          )
CERTIFICATE OF SERVICE Service has been effected on this date by personal delivery or first-class mail to the following:
                                    *** Marshall E. Miller, Esquire Chairman Atomic Safety & Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C. 20545 (2 copies)
                                  *****Dr. Cadet H. Hand, Jr.
Director Bodega Marine Laboratory University of California P. O. Box 247 Bodega Bay, California 94923
                                    ***Mr. Gustave A. Linenberger Atomic Safety & Licensing Board U. S. Nuclear Regulatory Commission      _
Washington, D. C. 20545
                                    **** Daniel Swanson, Esq.
                                    ****Stuart Treby, Esq.
Office of Executive Legal Director U. S. Nuclear Regulatory Commission    ,
Washington, D. C. 20545 (2 copies) l l
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      , _ , . _ _ . . . , . _ . .      _ . _ _ . . _ _}}

Latest revision as of 09:21, 16 December 2024

Proposed Partial Initial Decision,Authorizing Site Preparation Activities Per 10CFR50.10(e)(1).Certificate of Svc Encl
ML20070N330
Person / Time
Site: Clinch River
Issue date: 01/24/1983
From: Edgar G, Luck W
ENERGY, DEPT. OF
To:
References
NUDOCS 8301250477
Download: ML20070N330 (304)
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